China
Africa
Explore chapters and articles related to this topic
Fusion
Published in William J. Nuttall, Nuclear Renaissance, 2022
The stellarator concept is one of the oldest in fusion research, having been developed by Lyman Spitzer of Princeton University in the early 1950s [51]. The key to the stellarator concept is that the helical toroidal motion of the plasma ions and electrons is not driven by a transformer action but rather is provided in its entirety by the use of externally applied fields. In order to achieve such a task, the field coils themselves must generate a rotational transform and need therefore to be geometrically far more complex than those used in tokamaks. The windings of a stellarator ensure closed magnetic surfaces and hence low plasma losses. With no requirement for transformer action nor an applied plasma current, stellarators naturally operate in continuous steady-state mode. This has long been regarded as a key advantage of the concept for real power station applications. However, the lack of a transformer action eliminates the possibility of resistive heating within the stellarator plasma and in the early years, it seemed unlikely that sufficient temperatures could be achieved for efficient fusion to take place. Thankfully there have been significant advances over the years in plasma heating technology via microwaves and neutral beam injection. Previous attempts to introduce energy for plasma heating had relied on pulses of induced current, but these caused kink instabilities, eroding one of the key benefits of the stellarator technology [52]. The real innovation, however, that has made the stellarator a truly practical proposition has been the improvement in computer performance. Only in recent years, it has been possible to model the performance of complex stellarator magnet designs on a computer, and indeed computer-aided design has greatly improved design and build capabilities.
Wendelstein 7-X Near Real-Time Image Diagnostic System for Plasma-Facing Components Protection
Published in Fusion Science and Technology, 2018
A. Puig Sitjes, M. Jakubowski, A. Ali, P. Drewelow, V. Moncada, F. Pisano, T. T. Ngo, B. Cannas, J. M. Travere, G. Kocsis, T. Szepesi, T. Szabolics
Wendelstein 7-X (W7-X) is a drift optimized nuclear fusion device of stellarator type built in Greifswald by the Max-Planck-Institut für Plasmaphysik1 (IPP). Its main goal is to prove that the stellarator design is suitable for a future fusion power plant with steady-state operation. With plasmas of up to 30 min and with a steady-state heating power of 10 MW, a continuous real-time data acquisition, analysis, and control system is necessary in order to protect the plasma-facing components (PFCs) from overheating.
Nuclear Assessment to Support ARIES Power Plants and Next-Step Facilities: Emerging Challenges and Lessons Learned
Published in Fusion Science and Technology, 2018
Even though the stellarator concept is one of the first approaches proposed in the 1950s, very little in the way of conceptual design studies has been made compared to tokamaks. The first large-scale fully optimized stellarator experiment in the world, W7-X (Ref. 73), began operation in 2017. The initial encouraging results have generated substantial interest in the stellarator as potential fusion power plant and are likely to raise interest even further in the future. Since the 1980s, six large-scale stellarator power plants have been developed in the United States, Germany, and Japan. These studies varied in scope and depth and encompassed a broad range of configurations, offering steady-state, transient-free, nondisruptive operations with low recirculating power. Interest in the compact stellarator (CS) concept increased during the 2000s because of the remarkable advances in theory, experimental results, and construction techniques. These developments inspired designing the ARIES-CS (Refs. 22 and 23) since compactness was also promoted as an economic advantage. Unlike tokamaks, the stellarator coils are not equally spaced on the IB and OB for better plasma containment. While the magnetic geometry of tokamaks is intended to be entirely symmetric in the toroidal coordinate, the magnetic field components of the stellarator vary in all three coordinates, deviating from toroidal symmetry. As such, the stellarator physics advantages are offset by the more complex configurations, harder divertor designs, and challenging maintenance schemes. In the ARIES-CS (Fig. 20 and Table III) and most stellarator designs, the FW and surrounding in-vessel components conform to the plasma and deviate from the uniform toroidal shape to reduce the machine size. As such, the FW shape varies toroidally and poloidally, representing a challenging 3-D engineering problem and making the design of in-vessel components, fabrication, overall integration process, and maintenance scheme more complex compared to tokamaks.