Bulletin of the American Physical Society
49th Annual Meeting of the Division of Plasma Physics
Volume 52, Number 11
Monday–Friday, November 12–16, 2007; Orlando, Florida
Session TM5: Mini-conference on the First Microns of the First Wall: Mixed Materials Issues |
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Chair: Tom Rognlien, Lawrence Livermore National Laboratory Room: Rosen Centre Hotel Salon 11/12 |
Thursday, November 15, 2007 9:30AM - 9:50AM |
TM5.00001: Review of C-Be mixed material plasma experiments in PISCES-B J. Hanna, D. Nishijima, R.P. Doerner, M. Baldwin, K.R. Umstadter, R. Seraydarian, R. Hernandez The current ITER design employs a Be first wall, and a W divertor with C strike points. ITER also calls for severe heat loads on these plasma-facing components. A beryllium-seeded deuterium plasma is used in PISCES-B to investigate mixed-material erosion and redeposition properties of ITER relevant divertor materials. Ongoing experiments using C samples in these Be-seeded deuterium plasmas will be reviewed. Of specific interest is the formation of beryllium carbide on the sample surface that effectively mitigates C chemical erosion. Changes in hydrogen retention in the samples due to mixed-material layer formation will also be discussed. Effects of transient heating of the samples to simulate surface temperature excursions on the carbide layer formation and hydrogen retention will also be presented. [Preview Abstract] |
Thursday, November 15, 2007 9:50AM - 10:10AM |
TM5.00002: Mixed plasma species effects on Tungsten Matt Baldwin, Russ Doerner, Daisuke Nishijima, Yoshio Ueda The diverted reactor exhaust in confinement machines like ITER and DEMO will be intense-mixed plasmas of fusion (D, T, He) and wall species (Be, C, W, in ITER and W in DEMO), characterized by tremendous heat and particle fluxes. In both devices, the divertor walls are to be exposed to such plasma and must operate at high temperature for long durations. Tungsten, with its high-melting point and low-sputtering yield is currently viewed as the leading choice for divertor-wall material in this next generation class of fusion devices, and is supported by an enormous amount of work that has been done to examine its performance in hydrogen isotope plasmas. However, studies of the more realistic scenario, involving mixed species interactions, are considerably less. Current experiments on the PISCES-B device are focused on these issues. The formation of Be-W alloys, He induced nanoscopic morphology, and blistering, as well as mitigation influences on these effects caused by Be and C layer formation have all been observed. These results and the corresponding implications for ITER and DEMO will be presented. [Preview Abstract] |
Thursday, November 15, 2007 10:10AM - 10:30AM |
TM5.00003: Sputtering, impurity transport, and redeposition at the divertor and first wall Jeffrey N. Brooks These wall-related processes are critical for ITER and future fusion reactor plasma facing component surfaces. REDEP/WBC code-package full kinetic, 3-D Monte Carlo calculations (with typical input of UEDGE/DEGAS plasma edge parameters and TRIM-SP sputter yields/distributions) are used to study the sputtering erosion/redeposition process. A major issue is formation of mixed surface materials e.g., Be/C or Be/W, and resulting sputtering, thermomechanical, and tritium codeposition properties. Generally, divertor-sputtered tungsten is highly locally redeposited with essentially zero net sputter erosion and plasma contamination predicted. Beryllium and physically sputtered carbon travel farther, but are still confined to the near-surface divertor region. In contrast, chemically sputtered divertor carbon and wall-sputtered material of any type can travel much further, with implications for T/C and T/Be codeposition, and Be or W wall-to-divertor transport and mixing. [Preview Abstract] |
Thursday, November 15, 2007 10:30AM - 10:50AM |
TM5.00004: Hydrogenic Fuel Retention in Refractory Metals D.G. Whyte, B. Lipschultz, J. Irby, G.M. Wright High-Z refractory metals such as tungsten and molybdenum (Mo) are favored as plasma-facing components in burning plasma experiment to minimize hydrogenic (H) fuel retention, mainly due to their low H solubility ($\sim $1 appm). Fuel retention in Mo has been measured and modeled in the Mo-tile Alcator C-Mod tokamak, and DIONISOS a new facility that features simultaneous plasma bombardment and real-time retention diagnosis in the first 10 microns of the material. We find that high ion fluxes and the requirement for surface recombination into D2 in order to release the deuterium, leads to a type of ``pressure-induced'' trap formation $>$ 1{\%} concentration in the metals; i.e. much larger than the solubility. Exposure in tokamaks leads to temperature transients through plasma heating and neutron bombardment that also increase retention. High temperature drives D traps permeation into the Mo, but the sudden cooling of the material with removal of the plasma flux ``freezes'' the D deep in the Mo which can only be released by temperatures significantly higher than obtained during plasma exposure. Nuclear displacements by high-energy particles (neutrons, $\sim $MeV ions) also lead to damage sites that greatly enhance retention. These retention mechanisms occur in the bulk of the material, and are fundamentally different than co-deposition of D with surface films. Implications for burning plasma experiments will be discussed. [Preview Abstract] |
Thursday, November 15, 2007 10:50AM - 11:10AM |
TM5.00005: Effect of impurities on the thermo-oxidative removal of codeposits from DIII-D and JET divertor tiles A.A. Haasz, C. Tsui, J.W. Davis Results are presented for the thermo-oxidative removal of codeposits from DIII-D and JET divertor tiles. The DIII-D codeposits are relatively thin (1-2 $\mu $m) and contain B impurities (0-45{\%}), while the JET codeposits are thick (up to $\sim $250 $\mu $m) and contain up to $\sim $75{\%} Be. Erosion rates, D-removal rates and the remaining D content in the codeposits were measured as a function of (i) pressure (2.1-79 kPa), (ii) temperature (523-673 K), and (iii) oxidation time (15 min to 8 h). The DIII-D results show that for C-D codeposits with less than a few percent B, it is possible to remove $>$95{\%} of the D content in the codeposit in 15 minutes. The D-removal rate decreases with higher levels of B concentration. Our first results for the `thick' Be-containing JET codeposits suggest that the initial rate of D removal is much higher for these thicker codeposits than for the previously studied relatively `thin' (1-2 $\mu $m) DIII-D and JET deposits. This is despite the large Be concentration. Implications for ITER will be discussed. [Preview Abstract] |
Thursday, November 15, 2007 11:10AM - 11:30AM |
TM5.00006: Innovative Tokamak First Wall and Divertor Material Concepts C.P.C. Wong For ITER design, the design guidance is to apply a Be layer on the plasma facing chamber surface. When extrapolated to DEMO design, the Be layer will not be suitable due to radiation damage. Similarly, a carbon surface will not be suitable due to high physical and chemical erosion rates, radiation damage of the material and potential large retention of tritium. Unfortunately, the remaining commonly proposed material, tungsten (W), could suffer radiation damage from $\alpha $-charged particle implantation and experience blistering at the first wall and the formation of submicron fine structure at the divertor, which could result in W transport to the plasma core and severely limit the core performance. To resolve this potential impasse, an invention on the use of boron-infiltrated W-mesh surface is proposed to withstand ELMs and disruptions while retaining the capability of transmitting high-grade heat for power conversion. To make this concept work, in-situ boronization will be needed. Innovative first wall and divertor material concepts will be reviewed and initial development and identified requirements for the BW-mesh concept will be reported. [Preview Abstract] |
Thursday, November 15, 2007 11:30AM - 11:50AM |
TM5.00007: Evolution of Elemental Composition and Morphology in Fusion Reactor's First Wall Yong W. Kim Forcing of a multi-element alloy by a gradient field can modify the spatial profile of its elemental composition. The gradient field may be in the imposed temperature or the flux of impinging particles. In a fusion device, both scenarios apply. The consequences must be well understood because they change the thermal transport properties as well as the strength, corrosion and wear characteristics of the first wall materials. Given the large number of directions material evolution can take, new robust methods of near-surface composition analyses are needed. This paper presents a new measurement methodology and requisite instrumentation, which can provide measures of local elemental composition and transport properties simultaneously by time-resolved spectroscopy of laser-produced plasma (LPP) plume emissions from the specimen surfaces. The studies to date show that the composition profiles can be modified thermally in a reproducible manner; disparate thermal transport of constituent atoms can incur modifications of near-surface composition profiles.[Y.W. Kim, Int. J. Thermophysics \textbf{\textit{28, }}732 (2007)] Also, disparate fluxes of fuel particles, fusion products and impurities force the first walls in myriad ways. Repetitive application of the LPP analysis can resolve the near-surface composition profile as well as transport properties over several microns with depth resolutions to 20 nm. Work supported in part by NSF-DMR. [Preview Abstract] |
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