This paper summarizes the latest progress in the ITER blanket system design as it proceeds through its final design phase with the Final Design Review planned for Spring 2013. The blanket design is constrained by demanding and sometime conflicting design and interface requirements from the plasma and systems such as the vacuum vessel, in-vessel coils and blanket manifolds. This represents a major design challenge, which is highlighted in this paper with examples of design solutions to accommodate some of the key interface and integration requirements.
Abstract. This paper summarizes highlights of research results from the Alcator C-Mod tokamak covering the period 2006 through 2008. Active flow drive, using mode converted waves in the ion cyclotron range of frequencies (ICRF), has been observed for the first time in a tokamak plasma, using a mix of D and 3He ion species; toroidal and poloidal flows are driven near the location of the mode conversion layer. ICRF induced edge sheaths are implicated in both the erosion of thin boron coatings and the generation of metallic impurities. Lower Hybrid RF has been used for efficient current drive, current profile modification, and toroidal flow drive. In addition, LHRF has been used to modify the H-mode pedestal, increasing temperature, decreasing density, and lowering the pedestal collisionality. Studies of hydrogen isotope retention in solid metallic plasma facing components reveal significantly higher retention than expected from ex-situ laboratory studies; a model to explain the results, based on plasma/neutral induced lattice damage has been developed and tested. During gaspuff mitigation of disruptions, induced MHD causes the magnetic field to become stochastic, resulting in reduction of halo currents, spreading of plasma power loading, and loss of run-away electrons before they cause damage. Detailed pedestal rotation profile measurements have been used to infer ER profiles, and correlation with global H-mode confinement. An improved L-mode regime, obtained at q 95 ≤3 with ion drift away from the active x-point, shows very good confinement without a strong density pedestal, and no evidence of particle or impurity accumulation without the need for ELMs or any other edge density regulation mechanism.
The advanced limiter-divertor plasma-facing systems (ALPS) program was initiated in order to evaluate the potential for improved performance and lifetime for plasma-facing systems. The main goal of the program is to demonstrate the advantages of advanced limiter/divertor systems over conventional systems in terms of power density capability, component lifetime, and power conversion efficiency, while providing for safe operation and minimizing impurity concerns for the plasma. Most of the work to date has been applied to free surface liquids. A multi-disciplinary team from several institutions has been organized to address the key issues associated with these systems. The main performance goals for advanced limiters and divertors are a peak heat flux of \ 50 MW/m 2 , elimination of a lifetime limit for erosion, and the ability to extract useful heat at high power conversion efficiency ( 40%). The evaluation of various options is being conducted through a combination of laboratory experiments, www.elsevier.com/locate/fusengdes * Corresponding author. Tel.: + 1-630-2528673; fax: + 1-630-2525287. 1 To be presented at the Fifth International Symposium on Fusion Nuclear Technology. The submitted manuscript has been created by the University of Chicago as Operator of Argonne National Laboratory ('Argonne') under Contract No. W-31-109-ENG-38 with the US Department of Energy. The US Government retains for itself, and others acting on its behalf, a paid-up, nonexclusive, irrevocable world-wide license in said article to reproduce, prepare derivative works, distribute copies to the public, and perform publicly and display publicly, by or on behalf of the Government.0920-3796/00/$ -see front matter © 2000 Elsevier Science B.V. All rights reserved. PII: S 0 9 2 0 -3 7 9 6 ( 0 0 ) 0 0 3 8 5 -9 49-50 (2000) 127-134 128 modeling of key processes, and conceptual design studies. The current emphasis for the work is on the effects of free surface liquids on plasma edge performance. R.F. Mattas et al. / Fusion Engineering and Design
The US Enabling Technology Program in fusion is investigating the use of free flowing liquid surfaces facing the plasma. We have been studying the issues in integrating a liquid surface divertor into a configuration based upon an advanced tokamak, specifically the ARIES-RS configuration. The simplest form of such a divertor is to extend the flow of the liquid first wall into the divertor and thereby avoid introducing additional fluid streams. In this case, one can modify the flow above the divertor to enhance thermal mixing. For divertors with flowing liquid metals (or other electrically conductive fluids) MHD (magneto-hydrodynamics) effects are a major concern and can produce forces that redirect flow and suppress turbulence. An evaluation of Flibe (a molten salt) as a working fluid was done to assess a case in which the MHD forces could be largely neglected. Initial studies indicate that, for a tokamak with high power density, an integrated Flibe first wall and divertor does not seem workable. We have continued work with molten salts and replaced Flibe with Flinabe, a mixture of lithium and sodium fluorides, that has some potential because of its lower melting temperature. Sn and Sn-Li have also been considered, and the initial evaluations on heat removal with minimal plasma contamination show promise, although the complicated 3-D MHD flows cannot yet be fully modeled. Particle pumping in these design concepts is accomplished by conventional means (ports and pumps). However, trapping of hydrogen in these flowing liquids seems plausible and novel concepts for entrapping helium are also being studied. 1 Sandia is a multi-program laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the NNSA United States Department of Energy under Contract DE-AC04-94AL85000.
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