Implications of NSTX lithium results for magnetic fusion research

Lithium wall coating techniques have been experimentally explored on National Spherical Torus Experiment (NSTX) for the last five years. The lithium experimentation on NSTX started with a few milligrams of lithium injected into the plasma as pellets and it has evolved to a lithium evaporation system which can evaporate up to ∼100 g of lithium onto the lower divertor plates between lithium re-loadings. The unique feature of the lithium research program on NSTX is that it can investigate the effects of lithium in H-mode divertor plasmas. This lithium evaporation system thus far has produced many intriguing and potentially important results; the latest of these are summarized in a companion paper by H. Kugel. In this paper, we suggest possible implications and applications of the NSTX lithium results on the magnetic fusion research which include electron and global energy confinement improvements, MHD stability enhancement at high beta, edge localized mode (ELM) control, H-mode power threshold reduction, improvements in radio frequency heating and non-inductive plasma start-up performance, innovative divertor solutions and improved operational efficiency.     

Thermal control of the liquid lithium divertor for NSTX

The liquid lithium divertor (LLD) to be installed in NSTX has four toroidal panels, each a conical section inclined at 22° like the previous graphite divertor tiles. Each LLD panel is a copper plate clad with 0.25 mm of stainless steel (SS) and a surface layer of flame sprayed molybdenum (Mo) that will host lithium deposited from an evaporator. LITER (evaporators) already used in NSTX will be upgraded for the LLD. Each has twelve 500 W cartridge heaters with thermocouples, 16 other thermocouples, and a channel for helium cooling. During LLD experiments, the LLD will be heated so that the lithium is just above its melting temperature. The length of each shot will be preset to prevent excessive evaporation of lithium from the LLD. This duration depends on the heat load and is likely to be in the range of less than a second to several seconds. Careful thermal control of the LLD is important to maximize the shot times and to guide operation of the LLD. This paper describes the layout of the LLD, its expected thermal performance, the control system, and supporting experiments and analysis. A companion paper in this conference, “Physics design requirements for the national spherical torus experiment liquid lithium divertor,” provides other information.     

Recent progress in research on tungsten materials for nuclear fusion applications in Europe

The current magnetic confinement nuclear fusion power reactor concepts going beyond ITER are based on assumptions about the availability of materials with extreme mechanical, heat, and neutron load capacity. In Europe, the development of such structural and armour materials together with the necessary production, machining, and fabrication technologies is pursued within the EFDA long-term fusion materials programme. This paper reviews the progress of work within the programme in the area of tungsten and tungsten alloys. Results, conclusions, and future projections are summarized for each of the programme’s main subtopics, which are: (1) fabrication, (2) structural W materials, (3) W armour materials, and (4) materials science and modelling. It gives a detailed overview of the latest results on materials research, fabrication processes, joining options, high heat flux testing, plasticity studies, modelling, and validation experiments.     

A strategic framework for the magnetic fusion energy program

Fusion is recognized as a sufficiently abundant and environmentally attractive energy source to sustain industrial society in the 21st century and beyond. This paper outlines a strategic framework for the U.S. magnetic fusion program that builds substantially on the high-quality research and the strong scientific and technological basis that has been established during the past two decades.     

Portuguese research program on nuclear fusion

The Portuguese research program on nuclear fusion is presented. The experimental activity associated with the tokamak ISTTOK as well as the work carried out in the frame of international collaboration are summarized. Future plans are briefly described.     

End product economics and fusion research program priorities

It is shown that deuterium based fusion fuels and reactors based on them face severe technological disadvantages in comparison with fission based systems as power sources for central station electric power plants. The author postulates the most plausible deuterium based fusion reactor consistent with the physics of the fusion reaction itself and compares this reactor (called OMR-DT) with existing fission reactors. Since neutrons are the main problem in fusion, the author suggests that a great deal more effort should be given to the study of non-Maxwellian plasmas with the emphasis on neutron-free fuel cycles. The author also suggests that the deuterium based fusion driver may play its best role as a fissile fuel producer.     

Surface chemistry analysis of lithium conditioned NSTX graphite tiles correlated to plasma performance

Lithium wall conditioning in NSTX has resulted in reduced divertor recycling, improved energy confinement, and reduced frequency of edge-localized modes (ELMs), up to the point of complete ELM suppression. NSTX tiles were removed from the vessel following the 2008 campaign and subsequently analyzed using X-ray photoelectron spectroscopy as well as nuclear reaction ion beam analysis. In this paper we relate surface chemistry to deuterium retention/recycling, develop methods for cleaning of passivated NSTX tiles, and explore a method to effectively extract bound deuterium from lithiated graphite. Li–O–D and Li–C–D complexes characteristic of deuterium retention that form during NSTX operations are revealed by sputter cleaning and heating. Heating to ～850 °C desorbed all deuterium complexes observed in the O 1s and C 1s photoelectron energy ranges. Tile locations within approximately ±2.5 cm of the lower vertical/horizontal divertor corner appear to have unused LiO bonds that are not saturated with deuterium, whereas locations immediately outboard of this region indicate high deuterium recycling. X-ray photo electron spectra of a specific NSTX tile with wide ranging lithium coverage indicate that a minimum lithium dose, 100–500 nm equivalent thickness, is required for effective deuterium retention. This threshold is suspected to be highly sensitive to surface morphology. The present analysis may explain why plasma discharges in NSTX continue to benefit from lithium coating thickness beyond the divertor deuterium ion implantation depth, which is nominally <10 nm.     

Liquid lithium divertor system for fusion reactor

One of the most critical issues for the steady state fusion reactor is the heat flux in the divertor target. This paper proposes a liquid lithium divertor system to solve this problem. The proposed divertor system consists of a liquid lithium target, an evaporation chamber and a differential evacuation chamber. The heat coming from the fusion plasma along the divertor leg is removed by evaporation of lithium. The lithium vapor is condensed on the wall and is circulated with a pump. The coolant temperature for the wall is high enough to drive a power generator. Narrow slits along the divertor leg and the differential evacuation chamber reduce leakage of lithium vapor to the plasma chamber. A preliminary estimation predicts that the lithium ion density in the core plasma is lower than the plasma density.     

Review of the magnetic fusion program by the 1986 ERAB Fusion Panel

The 1986 ERAB Fusion Panel finds that fusion energy continues to be an attractive energy source with great potential for the future, and that the magnetic fusion program continues to make substantial technical progress. In addition, fusion research advances plasma physics, a sophisticated and useful branch of applied science, as well as technologies important to industry and defense. These factors fully justify the substantial expenditures by the Department of Energy in fusion research and development (R&D). The Panel endorses the overall program direction, strategy, and plans, and recognizes the importance and timeliness of proceeding with a burning plasma experiment, such as the proposed Compact Ignition Tokamak (CIT) experiment.Presented to the Magnetic Fusion Advisory Committee (La Jolla, California, December 4, 1986)     

Recent progress on new treatment research project at HIMAC

On the basis of more than 10 years of experience with HIMAC, we proposed and designed a new treatment research facility, as an extension of the existing facility, for the further development of radiotherapy with HIMAC. The new facility employs a 3D rescanning technique developed at NIRS. Its building construction was completed in March 2010 and the devices are being installed. A beam commissioning and a pre-clinical study will be initiated from this October. Here, we report recent progress for this new treatment facility project at HIMAC.     

SiC MOSFET单粒子效应研究现状

伴随核能与空间技术的快速发展,SiC MOSFET等高压大功率器件的应用不断增加,其因环境中的高能粒子辐射所引起的单粒子效应问题（如单粒子烧毁、单粒子栅击穿等）也逐渐凸显。为全面深入认识该问题,首先,论证了SiC MOSFET的优势特性,及其在辐射应用中面临的关键问题。然后,整理了目前国内外关于SiC MOSFET单粒子效应的模拟计算、辐照实验及相应研究成果,总结了在SiC MOSFET单粒子效应研究中的主要关注点,并分析了SiC MOSFET单粒子效应敏感性较高的原因。最后,基于目前SiC MOSFET单粒子效应研究中仍存在的问题,展望了未来可重点关注的研究方向。通过系统总结国内外SiC MOSFET单粒子效应研究进展,希望能为研究SiC MOSFET单粒子效应物理机制以及改进其抗单粒子效应加固技术提供有价值的参考。     

A simple apparatus for the injection of lithium aerosol into the scrape-off layer of fusion research devices

A simple device has been developed to deposit elemental lithium onto plasma facing components in the National Spherical Torus Experiment. Deposition is accomplished by dropping lithium powder into the plasma column. Once introduced, lithium particles quickly become entrained in scrape-off layer flow as an evaporating aerosol. Particles are delivered through a small central aperture in a computer-controlled resonating piezoelectric disk on which the powder is supported. The device has been used to deposit lithium both during discharges as well as prior to plasma breakdown. Clear improvements to plasma performance have been demonstrated. The use of this apparatus provides flexibility in the amount and timing of lithium deposition and, therefore, may benefit future fusion research devices.     

A review of plasma-wall boundary effects on core confinement and lithium applications to boundary-controlled magnetic fusion experiments

Reduced particle recycling in the edge region is often considered to be a necessary condition to achieve high-performance confinement in magnetic fusion experiments. However, it is also true that whenever or however core confinement improves, as a result edge recycling is reduced. To provide a logical interpretation of this circular cause-and-consequence relation, reviewed in this paper are some of the important data demonstrating plasma-wall boundary effects on core confinement and also those evaluating innovative wall concepts with lithium applied to sustain optimized boundary conditions for the operation of steady-state fusion power reactors.     

Heavy-section steel technology program: Recent developments in crack initiation and arrest research

Technology for the analysis of crack initiation and arrest is central to the reactor pressure vessel fracture-margin-assessment process. Regulatory procedures for nuclear plants utilize this technology to assure the retention of adequate fracture-prevention margins throughout the plant operating license period. As nuclear plants age and regulatory procedures dictate that fracture- margin assessments be performed, interest in the fracture-mechanics technology incorporated into those procedures has heightened. This has led to proposals from a number of sources for development and refinement of the underlying crack-initiation and arrest-analysis technology. This paper presents an overview of ongoing Heavy-Section Steel Technology (HSST) Program research aimed at refining the fracture toughness data used in the analysis of fracture margins under pressurized-thermal-shock loading conditions.