Showing papers in "Fusion Science and Technology in 2008"

Recent advances on hydrogen retention in ITER's plasma-facing materials: Beryllium, carbon and tungsten

Abstract: Management of tritium inventory remains one of the grand challenges in the development of fusion energy, and the choice of plasma-facing materials is a key factor for in-vessel tritium retention. The Atomic and Molecular Data Unit of the International Atomic Energy Agency organized a Coordinated Research Project (CRP) on the overall topic of tritium inventory in fusion reactors during the period 2001-2006. This dealt with hydrogenic retention in ITER’s plasma-facing materials – Be, C, and W – and in compounds (mixed materials) of these elements as well as tritium removal techniques. The results of the CRP are summarized in this paper together with recommendations for ITER. Basic parameters of diffusivity, solubility, and trapping in Be, C, and W are reviewed. For Be, the development of open porosity can account for transient hydrogenic pumping, but long-term retention will be dominated by codeposition. Codeposition is also the dominant retention mechanism for carbon and remains a serious concern f...

The aries-cs compact stellarator fusion power plant

Abstract: An integrated study of compact stellarator power plants, ARIES-CS, has been conducted to explore attractive compact stellarator configurations and to define key research and development (R&D) areas The large size and mass predicted by earlier stellarator power plant studies had led to cost projections much higher than those of the advanced tokamak power plant As such, the first major goal of the ARIES-CS research was to investigate if stellarator power plants can be made to be comparable in size to advanced tokamak variants while maintaining desirable stellarator properties As stellarator fusion core components would have complex shapes and geometry, the second major goal of the ARIES-CS study was to understand and quantify, as much as possible, the impact of the complex shape and geometry of fusion core components This paper focuses on the directions we pursued to optimize the compact stellarator as a fusion power plant, summarizes the major findings from the study, highlights the key design aspects and constraints associated with a compact stellarator, and identifies the major issues to help guide future R&D

Chapter 7: Tomography Diagnostics: Bolometry and Soft-X-Ray Detection

Abstract: This chapter reviews multichannel broadband measurement of the soft-X-ray radiation and total radiation in magnetically confined fusion plasma experiments Common detector types used (including bolometers), details of their application, and interpretation of their measurements are described An introduction is given to the application of computed tomography methods in the mathematical reconstruction of emission profiles from multiple (approximately) line-integral measurements, taking into account the specific circumstances common in magnetically confined fusion plasma experiments Although the emphasis is on two-dimensional tomography of poloidal cross sections, the applications of Abel inversion, three-dimensional tomography, vector tomography, and other specific methods are briefly discussed Several examples of the application and the plasma parameters that can be derived are given

Chapter 12: Generic Diagnostic Issues for a Burning Plasma Experiment

Abstract: In this chapter, we consider generic issues affecting the implementation of diagnostics in a burning plasma experiment (BPX). These are, directly or indirectly, caused by the radiation environment....

Design of Circular Corrugated Waveguides to Transmit Millimeter Waves at ITER

Abstract: Low-loss circular waveguides will be needed for a large number of millimeter-wave transmission lines on ITER, including those transmitting electron cyclotron power and diagnostic signals. In order to provide low-loss transmission, the waveguides need to be several wavelengths in diameter. Corrugating the walls reduces the loss further not only in straight runs but also at bends, and makes the waveguide robust against small deformations. We present results of theoretical calculations showing that these properties can be maintained over very wide bandwidths suitable for ITER applications. The computer code used to make these calculations is based on a space-harmonic analysis of the fields. Measurements on waveguides are described that validate the theory for corrugated waveguides semiquantitatively. Tolerances on the corrugation geometry, waveguide bore, waveguide junctions, input Gaussian beam alignment, and waveguide support alignment are discussed. It is shown that the low-loss properties of corr...

Chapter 2: Magnetic Diagnostics

Abstract: Magnetic diagnostics are essential for the operation and understanding of a magnetic fusion device. Magnetic data are used in real time to measure and control the current, shape, and position of the discharge; the thermal energy of the plasma; the confining magnetic field; and the currents in the magnet coils. Equilibrium reconstructions based on magnetic data yield the magnetic geometry of the plasma, providing the coordinates for interpretation of all other diagnostic measurements. Magnetic measurements also provide input for the analysis and feedback control of magnetohydrodynamic (MHD) instabilities. This review focuses on the inductive loops and Hall effect probes that are used in nearly all present devices. We describe the principles of magnetic diagnostics and discuss issues related to their practical implementation. The interpretation of magnetic measurements for equilibrium reconstruction and for identification of MHD instabilities are summarized. Magnetic diagnostics based on inductive m...

Chapter 5: Passive Spectroscopic Diagnostics for Magnetically Confined Fusion Plasmas

Abstract: Spectroscopy of radiation emitted by impurities and hydrogen isotopes plays an important role in the study of magnetically confined fusion plasmas, both in determining the effects of impurities on ...

Chapter 3: Microwave Diagnostics

Abstract: Microwave-based diagnostics have found broad application in magnetic fusion plasma diagnostics and are expected to be widely employed in future burning plasma experiments (BPXs). Most of these tech...

Edge Localized Modes in Tokamaks

Abstract: As one increases the heating power in a tokamak beyond a threshold, the confinement undergoes a bifurcation, with a dramatic increase in the confinement time by a factor ˜2. This improved confineme...

Engineering design and analysis of the aries-cs power plant

Abstract: The ARIES-CS team has concluded an integrated study of a compact stellarator power plant, involving physics and engineering design optimization. Key engineering considerations include the size of t...

Chapter 6: Active Spectroscopy

Abstract: In this chapter we discuss the various diagnostic techniques in which active spectroscopy plays a role. The use of an injected neutral beam - either a dedicated diagnostic beam or the main heating beams - to localize and enhance the spectroscopic measurements described in Chap. 5 has been exploited for a number of key physics measurements, in particular detailed profile information on ion parameters, the radial electric field, plasma current density, and turbulent transport. The ability to make these detailed measurements has been a key element in the development of improved plasma performance. The neutral beam techniques have been extended by the use of such beam analogs as laser beams, gas puffs, and pellet injection for specific measurements. In each case we describe the general principle behind the measurement and include several successful examples of their implementation, briefly touching on some of the more important physics results. We conclude with a few remarks about the relevance and re...

Physics Design for ARIES-CS

Abstract: Novel stellarator configurations have been developed for ARIES-CS. These configurations are optimized to provide good plasma confinement and flux surface integrity at high beta. Modular coils have been designed for them in which the space needed for the breeding blanket and radiation shielding was specifically targeted such that reactors generating GW electrical powers would require only moderate major radii (<10 m). These configurations are quasi-axially symmetric in the magnetic field topology and have small number of field periods (≤3) and low aspect ratios (≤6). The baseline design chosen for detailed systems and power plant studies has 3 field periods, aspect ratio 4.5 and major radius 7.5 m operating at β~6.5% to yield 1 GW electric power. The shaping of the plasma accounts for ≥75% of the rotational transform. The effective helical ripples are very small (< 0.6% everywhere) and the energy loss of alpha particles is calculated to be ≤5% when operating in high density regimes. An interesting feature in this configuration is that instead of minimizing all residues in the magnetic spectrum, we preferentially retained a small amount of the non-axisymmetric mirror field. The presence of this mirror and its associated helical field alters the ripple distribution, resulting in the reduced ripple-trapped loss of alpha particles despite the long connection length in a tokamak-like field structure. Additionally, we discuss two other potentially attractive classes of configurations, both quasi-axisymmetric: one with only two field periods, very low aspect ratios (~2.5), and less complex coils, and the other with the plasma shaping designed to produce low shear rotational transform so as to assure the robustness and integrity of flux surfaces when operating at high β.

Designing aries-cs compact radial build and nuclear system: neutronics, shielding, and activation

Abstract: Within the ARIES-CS project, design activities have focused on developing the first compact device that enhances the attractiveness of the stellarator as a power plant. The objectives of this paper...

Transmission Line Technology for Electron Cyclotron Heating

Abstract: We briefly review the history of transmission line technology for electron cyclotron heating (ECH) applications and then survey the major developments over the past few years. These developments are grouped by function. Papers in this special issue are highlighted. We concentrate on the transmission from a matching optics unit near a gyrotron microwave source to the location of a vacuum window near the plasma, without review of window and launcher technology. Prospects for components handling more than 1-MW continuous wave are reviewed. While both waveguide and free-space propagation are considered, a greater emphasis is placed on corrugated waveguide components in preparation for the ITER ECH system.

Disproportionation Characteristics of a Zirconium-Cobalt Hydride Bed under ITER Operating Conditions

Abstract: Quantitative assessment of a disproportionation in the ZrCo-hydrogen system under ITER-relevant operating conditions was performed by means of experimental tests and a theoretical calculation. In the static temperature experiments with equilibrium hydrogen pressures, a 10% disproportionation of ZrCoHx (x = 2.0 and 2.5) was observed in 5.5 h at 415° (~78 kPa), 9 h at 400° (~72 kPa), 172 h at 380° (~51 kPa), and 1626 h at 350° (~28 kPa). An experimental formula [log τ = 17 268/T (K) - 25.814, where τ is the reaction time (day) of 10% disproportionation] was derived from these experiments. Experiments with a temperature cycling of up to 125 cycles (from room temperature to 350 to 360°) proved that no enhancement of a disproportionation occurs in the ZrCoHx (1.7 < x ≤ 2.0). Typical operation conditions of the ZrCo hydride bed for the D-T gas storage delivery system were proposed based on the ITER FDR 2000 plasma operation scenarios. The disproportionation rate estimated conservatively by the theoretic...

Safety Assessment of the ARIES Compact Stellarator Design

Abstract: ARIES-CS is a 1000 MW(electric) compact stellarator conceptual fusion power plant design. This power plant design contains many innovative features to improve the physics, engineering, and safety performance of the stellarator concept. ARIES-CS utilizes a dualcooled lead lithium blanket that employs low-activation ferritic steel as a structural material, with the first wall cooled by helium and the breeding zone self-cooled by flowing lead lithium. In this paper we examine the safety and environmental performance of ARIES-CS by reporting radiological inventories, decay heat, and radioactive waste management options and by examining the response of ARIES-CS to accident conditions. These accidents include conventional loss of coolant and loss of flow events, an ex-vessel loss of coolant event, and an in-vessel loss of coolant with bypass event that mobilizes in-vessel radioactive inventories (e.g., tritium and erosion dust from plasma-facing components). Our analyses demonstrate that the decay heat can be safely removed fromARIES-CSandthefacilitycanmeettheno-evacuation requirement.

Fusion product diagnostics

Abstract: Fusion product diagnostics can be used to determine a fusion reaction rate, which indicates how close the plasma is to the ultimate goal of making a power plant based on nuclear fusion. However, th ...

Overview of the National Ignition Facility

Abstract: The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory will be the world's largest and most powerful laser system for inertial confinement fusion (ICF) and experiments studying high energy density (HED) science. NIF is a 192 beam Nd-glass laser facility that will produce 1.8 MJ, 500 TW of ultraviolet light making it over fifty times more energetic than present ICF facilities. The NIF Project began in 1995 and is scheduled for completion in 2009. Ignition experiments on NIF, which will use tritium, are scheduled to begin in 2010. Tritium will arrive at the facility in individual target assemblies. The assemblies will be mounted to the Cryogenic TARget POSitioner (TARPOS), which provides the cryogenic cooling systems necessary to complete the formation of the ignition target's fuel ice layer. It also provides the positioning system that transports and holds the target at the center of the NIF chamber during a shot. After a shot, unburned tritium will be captured by the cryopumps. Upon regeneration, the cryopump effluent will be directed to the Tritium Processing System, part of NIF's. Personnel and Environmental Protection Systems. These systems also include, local contamination control systems, area and stack tritium monitoring systems, a decontamination area, andmore » waste packaging and characterization capability. This equipment will be used along with standard contamination control practices to manage the tritium hazard to workers and to limit releases to the environment to negligibly small amounts.« less

Recent progresses in Tritium radioecology and dosimetry

Abstract: In this paper, some aspects of recent progress in tritium radioecology and dosimetry are presented, with emphasis on atmospheric releases to terrestrial ecosystems. The processes involved in tritium transfer through the environment are discussed, together with the current status of environmental tritium models. Topics include the deposition and reemission of HT and HTO, models for the assessment of routine and accidental HTO emissions, a new approach to modeling the dynamics of tritium in mammals, the dose consequences of tritium releases and aspects of human dosimetry. The need for additional experimental data is identified, together with the attributes that would be desirable in the next generation of tritium codes.

Chapter 3: ELMy H-Mode Operation in JET

Abstract: A wide range of studies on JET have contributed greatly to the development of the ELMy H-mode as a high-performance scenario for fusion devices and to the understanding of the physical processes that underlie it. Development has focused on the production of a high-performance, high-density, stationary scenario suited to deuterium-tritium operation and with small edge energy loads. Physics studies have made strong progress in the understanding of the L-H threshold, energy confinement, pedestal physics, and edge-localized mode behavior. A strong focus of this work has been providing a basis for extrapolation to future machines, such as ITER, for which, as the largest existing tokamak, JET has been of particular importance.

Development of Water Detritiation Facility for JET

Abstract: This paper presents results from the concept evaluation, experimental trials, and design of a water detritiation facility (WDF) for the JET fusion machine. The design is based on the combined elect...

Chapter 10 - Core transport studies in jet

Abstract: Reference CRPP-ARTICLE-2009-033View record in Web of Science Record created on 2009-01-28, modified on 2017-05-12

First Tritium Results of the KATRIN Test Experiment Trap

Abstract: The TRAP experiment (TRitium Argon frost Pump) has been built at the Tritium Laboratory Karlsruhe (TLK) as a test rig for the Cryogenic Pumping Section (CPS) of the KArlsruhe TRItium Neutrino Experiment (KATRIN). TRAP employs a heterogeneous layer of pre-condensed argon to adsorb hydrogen isotopes at ∼ 4.2 K. This paper presents results obtained in the first three tritium experiments with TRAP.

Plasma-Wall Interaction in Magnetically Confined Fusion Plasmas

Abstract: The control of wall loads in fusion devices, in particular with respect to the life time limitations of wall components due to material erosion and migration, will be decisive for the realisation o...

Chapter 5 burning plasma studies at jet

Abstract: Studies establishing key phenomena and developing diagnostics for energetic particle physics, which are essentialforthenextstepburningplasmaexperimentssuch as the International Thermonuclear Experimental Reactor (ITER), have been performed at the Joint European Torus (JET). Experiments have demonstrated clear selfheating of deuterium-tritium (D-T) plasma by alpha particlesasamaximuminelectrontemperatureatanoptimum mixture of 60 6 20% tritium. The change in electron temperature produced by alpha heating, DTe(0) 1.3 6 0.23keV,wasasexpectedfromclassicalheating,whereas the heating of thermal ions was higher than expected from reference deuterium discharges. Alfven eigenmodes were stable in the highest fusion performance D-T plasmas, in agreement with the modeling. Systematic studies on the existence and properties of Alfven eigenmodes with external antenna driving and detecting Alfven eigenmodes are presented. The formation of fuel ion tails due to alpha-particle knock-on effects is described as derived from neutral particle analyzer and neutron emission spectrometry in D-T experiments. The gamma-ray diagnostics are shown to measure profiles and energy distribution functions of high-energy ions and alpha particles.Time-andspace-resolvedgamma-rayimagesdemonstrated for the first time the possibility of measuring several types of energetic ions simultaneously. The novel technique of detecting unstable Alfven eigenmodes with interferometry is found to be superior in detecting corelocalized Alfven eigenmodes.

Physics studies with the additional heating systems in JET

Abstract: The physics studies of the three "heating" systems that are installed on JET are reviewed. Results from the beginning of JET up to now are presented with some emphasis on the more recent ones. The systems were used not only for heating, where JET has laid the groundwork to qualify them for heating the next generation of machines to ignition, but also increasingly as tools to control the plasma. This role, already exploited on JET, will become more and more important in the next machine, since main heating will be provided by the alpha-particle heating and one will have to rely on the heating systems to control the plasma during the burn.

A Decade of Tritium Technology Development and Operation at the Tritium Laboratory Karlsruhe

Abstract: The Tritium Laboratory Karlsruhe (TLK) has been designed to handle relevant amounts of tritium for the development of tritium technology for fusion reactors. This paper describes the tritium technology development and experience gained during the upgrade of facilities, interventions, replacement of failed components and operation of the TLK since its commissioning with tritium in 1994. (authors)

Critical Design Issues of the ITER ECH Front Steering Upper Launcher

Abstract: Reference CRPP-CONF-2009-017View record in Web of Science Record created on 2009-01-27, modified on 2017-05-12

ARIES-CS Maintenance System Definition and Analysis

Abstract: The goal of the ARIES compact stellarator is to define and assess a stellarator-based fusion power plant to provide electrical power as competitively as possible by balancing performance, cost, and...