Nonlinear Burn Condition and Kinetic Profile Control in Tokamak Fusion Reactors

One of the most promising devices for realizing power production through nuclear fusion is the tokamak. In order to maximize performance, it is preferable that tokamaks achieve operating scenarios characterized by good plasma confinement, improved magnetohydrodynamic stability, and a largely non-inductively driven plasma current. Such scenarios could enable steady-state reactor operation with high fusion gain, the ratio of fusion power produced to the external heating power needed to sustain reactions. There are many experimental tokamaks around the world, each exploring different facets of plasma physics and fusion technology. These experiments have reached the point where the power released from fusion is nearly equal to the power input required to heat the plasma. The next experimental step is ITER, which aims to reach a fusion gain exceeding ten for short pulses, and to sustain a gain of five for longer pulses (around 1000 s). In order for ITER to be a success, several challenging control engineering problems must be addressed.Among these challenges is to precisely regulate the plasma density and temperature, or burn condition. Due to the nonlinear and coupled dynamics of the system, modulation of the burn condition (either during ramp-up/shut-down or in response to changing power demands) without a well designed control scheme could result in undesirable transient performance. Feedback control will also be necessary for responding to unexpected changes in plasma confinement, impurity content, or other parameters, which could significantly alter the burn condition during operation. Furthermore, although stable operating points exist for most confinement scalings, certain conditions can lead to thermal instabilities. Such instabilities can either lead to quenching or a thermal excursion in which the system moves to a higher temperature equilibrium point. In any of these situations, disruptive plasma instabilities could be triggered, stopping operation and potentially causing damage to the confinement vessel.In this work, the problem of burn condition control is addressed through the design of a nonlinear control law guaranteeing stability of desired equilibria. Multiple actuation methods, including auxiliary heating, isotopic fueling, and impurity injection, are used to ensure the burn condition is regulated even when actuators saturate. An adaptive control scheme is used to handle model uncertainty, and an online optimization scheme is proposed to ensure that the plasma is driven to an operating point that minimizes an arbitrary cost function. Due to the possible limited availability of diagnostic systems in ITER and future reactors, an output feedback control scheme is also proposed that combines the nonlinear controller with an observer that estimates the states of the burning plasma system based on available measurements. Finally, the control scheme is tested using the integrated modeling code METIS.The control of spatial profiles of parameters, including current, density, and temperature, is also an important challenge in fusion research, due to their effect on MHD stability, non-inductive current drive, and fusion power. In this work, the problem of kinetic profile control in burning plasmas is addressed through a nonlinear boundary feedback control law designed using a technique called backstepping. A novel implementation of the backstepping technique is used that enables the use of both boundary and interior actuation. The backstepping technique is then applied to the problem of current profile control in both low-confinement and high-confinement mode discharges in the DIII-D tokamak based on a first-principles-driven model of the current profile evolution. Both designs are demonstrated in simulations and experimental tests

Integrated Control in Tokamaks using Nonlinear Robust Techniques and Actuator Sharing Strategies

Tokamaks are devices whose final purpose is obtaining energy by means of nuclear fusion reactions. To achieve such purpose, a gas is injected into the tokamak\u27s torus-shaped chamber and heated up to extremely high temperatures, giving birth to a plasma. When the necessary conditions of temperature, density, and confinement time are achieved, virtually inexhaustible energy can be produced in a tokamak. The main contribution of this dissertation is the development of integrated control strategies for tokamak plasmas. The development of integrated control architectures is necessary for tokamaks to become efficient and commercially competitive power plants. Because a tokamak plasma is a highly nonlinear, coupled dynamical system, the great diversity of complex control problems that coexist in a tokamak are indeed closely interrelated. However, this variety of control problems must be tackled by means of a limited number of actuators. A functional design for integrated tokamak-control architectures should employ multi-input multi-output controllers to simultaneously regulate as many plasma variables as possible with the available actuators. Supervisory and exception handling systems that monitor the plasma state arise as a necessity to ensure a safe tokamak-operation. Finally, actuator sharing and management capabilities should also exist in order to utilize the available actuators in an optimal way. Various control problems are tackled in this dissertation, including kinetic, magnetic, and instability control problems. Control-oriented, physics-based models that characterize some specific aspects of the plasma dynamics have been employed to develop new control-oriented simulation codes and integrated-control solutions that employ nonlinear, robust control techniques and optimization-based actuator-management strategies. Some of those control solutions have been experimentally tested in the DIII-D tokamak

Development and Validation of a Computational Tool for Fusion Reactors\u27 System Analysis

On the roadmap to fusion energy the development and the operation of a demonstration power plant (DEMO) is the next step after ITER, a key facility currently devoted to the exploration of the physics aspects for self-sustained fusion plasmas with sizes and fusion power comparable to those attended in fusion power plants (FPP). Fusion systems codes are essential computational tools aimed to simulate the physics and the engineering features of a FPP. The main objective of a system code is to find one (or more) reactor configurations which simultaneously comply with physics operational limits, engineering constraints and net electric output requirements. As such simulation tools need to scope many design solutions over a large parameter phase space, they rely on rather basic physics and engineering models (mostly at zero or one-dimensional level) and on a relatively large number of input specifications. Within the conceptual design of a FPP, systems codes are interfaced to the detailed transport codes and engineering platforms, which operate in much larger time scales. To fill the gap between systems and the detailed transport and engineering codes the high-fidelity system/design tool MIRA (Modular Integrated Reactor Analysis) has been developed. MIRA relies on a modular structure and provides a refined FPP system analysis, with the primary goal of generating a more robust plant baseline. It incorporates into a unique computing environment a mathematical algorithm for the utmost tokamak fusion problems, including two-dimensional plasma magnetic equilibrium and core physics, transport of neutron and photon radiations emitted from the plasma and electromagnetic and engineering characterization of the toroidal field (TF) and poloidal field (PF) field coil systems. Most of the implemented modules rely on higher spatial resolution compared to presently available system codes, such as PROCESS. The multiphysics MIRA approach has been applied to the DEMO 2015 baseline, generated by means of the PROCESS system code. The analysis has been carried out by taking an identical set of input assumptions and requirements (e.g. same fusion power, major radius and aspect ratio) and observing the response on certain figures of merit. This verification study has featured the violation of some constraining conditions imposed on plasma safety factor, TF ripple and plasma burn time. The DEMO 2015 baseline has been found not in line with all the imposed requirements and constraints, hence necessitates a set of active measures on some of the input parameters. Such measures have been reported in form of parameter scans, where three variables have been identified, such as plasma internal inductance, blanket breeding zone inboard thickness and vacuum vessel/TF coil gap radial outboard width. The addressed sensitivity analyses have shown non-trivial inter-parametric dependencies, never explored in fusion system analyses. For instance, large influences of the plasma internal inductance on safety factor, plasma shape, density and temperature features, peak divertor flux and plasma burn time have been observed. Moreover, an optimal overall breeding blanket + TF coil inboard width has been observed with respect to the maximization of the plasma burn time, representing a meeting point between neutronic tritium breeding and technological limits in central solenoid and TF coils superconducting cables. These outcomes have inspired important changes in the way of designing a tokamak reactor like DEMO, where more extended analyses of the key physics and engineering aspects of the reactor can speed up and improve the design process of a FPP

Development of a concept and basis for the DEMO diagnostic and control system

An initial concept for the plasma diagnostic and control (D&C) system has been developed as part of European studies towards the development of a demonstration tokamak fusion reactor (DEMO). The main objective is to develop a feasible, integrated concept design of the DEMO D&C system that can provide reliable plasma control and high performance (electricity output) over extended periods of operation. While the fusion power is maximized when operating near to the operational limits of the tokamak, the reliability of operation typically improves when choosing parameters significantly distant from these limits. In addition to these conflicting requirements, the D&C development has to cope with strong adverse effects acting on all in vessel components on DEMO (harsh neutron environment, particle fluxes, temperatures, electromagnetic forces, etc.). Moreover, space allocation and plasma access are constrained by the needs for first wall integrity and optimization of tritium breeding. Taking into account these boundary conditions, the main DEMO plasma control issues have been formulated, and a list of diagnostic systems and channels needed for plasma control has been developed, which were selected for their robustness and the required coverage of control issues. For a validation and refinement of this concept, simulation tools are being refined and applied for equilibrium, kinetic and mode control studies

Thermal-hydraulic optimization of a proposed EU-DEMO hydrogen passive removal system

As the R&D of magnetic fusion power demonstrating plants are approaching important steps toward concept designs, analysts are working parallelly on the safety assessment of such concepts to identify any potential risk. One of the safety concerns involves the confinement of radioactive substances during normal operation and accidental conditions. Several accident sequences inside the tokamak vacuum vessel or pressure suppression systems are characterized by the risk of hydrogen buildup and subsequent ignition that could threaten the structural confinement integrity. In the Safety and Environment work package of the EUROfusion consortium, possible approaches to mitigate the hydrogen explosion risk are under investigation. One of the exploratory solutions is based on limiting the hydrogen concentration that could reach flammable gas mixture conditions and using Passive Autocatalytic Recombiners (PARs) installed into the atmosphere of the pressure suppression systems tanks to recombine hydrogen. This paper examines the theoretical effectiveness of the PARs intervention during an in-vessel loss of coolant accident without the intervention of the decay heat removal system for the Water-Cooled Lithium Lead (WCLL) concept of EU-DEMO, using an optimization methodology. The involved systems have been modelled in MELCOR to estimate the PARs recombination capability as a function of the thermal-hydraulic parameters of the suppression tanks. Furthermore, the optimizer entity of the RAVEN tool is applied to perform optimization studies on the hydrogen recombination system design parameters. The goal is to explore the geometrical and thermal-hydraulic parameters that maximize the capability of the hydrogen removal system for the WCLL concept