Publications

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The low solubility of hydrogen in tungsten leads to the growth of near-surface hydrogen precipitates during high-flux plasma exposure, strongly affecting migration and trapping in the material. We have developed a continuum-scale model of precipitate growth that leverages existing techniques for simulating the evolution of 3He gas bubbles in metal tritides. The present approach focuses on bubble growth by dislocation loop punching, assuming a diffusing flux to nucleation sites that arises from ion implantation. The bubble size is dictated by internal hydrogen pressure, the mechanical properties of the material, as well as local stresses. In this article, we investigate the conditions required for bubble growth. Recent focused ion beam (FIB) profiling studies that reveal the sub-surface damage structure provide an experimental database for comparison with the modeling results. © 2010 Elsevier B.V. All rights reserved.

The tungsten ITER divertor will be operated at temperatures above 1000 K. Most of the laboratory experiments on hydrogen isotope retention in tungsten have been performed at lower temperatures where the hydrogen is retained as both atoms and molecules. At higher temperatures, atomic trapping plays a smaller role. The purpose of this paper is to see if hydrogen is trapped at internal voids at elevated temperatures, and to see if gas-filled cavities can be formed at high fluences. Additionally, this paper examines the effect of helium bubbles and radiation damage on trapping. © 2010 Elsevier B.V. All rights reserved.

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Chemical vapor deposited (CVD) silicon carbide (SiC) has superior resistance to tritium permeation even after irradiation. Prior work has shown Ultrametfoam to be forgiving when bonded to substrates with large CTE differences. The technical objectives are: (1) Evaluate foams of vanadium, niobium and molybdenum metals and SiC for CTE mitigation between a dense SiC barrier and steel structure; (2) Thermostructural modeling of SiC TPB/Ultramet foam/ferritic steel architecture; (3) Evaluate deuterium permeation of chemical vapor deposited (CVD) SiC; (4) D testing involved construction of a new higher temperature (> 1000 C) permeation testing system and development of improved sealing techniques; (5) Fabricate prototype tube similar to that shown with dimensions of 7cm {theta} and 35cm long; and (6) Tritium and hermeticity testing of prototype tube.

Under appropriate conditions, uranium will form a hydride phase when exposed to molecular hydrogen. This makes it quite valuable for a variety of applications within the nuclear industry, particularly as a storage medium for tritium. However, some aspects of the U+H system have been characterized much less extensively than other common metal hydrides (particularly Pd+H), likely due to radiological concerns associated with handling. To assess the present understanding, we review the existing literature database for the uranium hydride system in this report and identify gaps in the existing knowledge. Four major areas are emphasized: {sup 3}He release from uranium tritides, the effects of surface contamination on H uptake, the kinetics of the hydride phase formation, and the thermal desorption properties. Our review of these areas is then used to outline potential avenues of future research.

Overview of Plasma Materials Interaction (PMI) activities are: (1) Hydrogen diffusion and trapping in metals - (a) Growth of hydrogen precipitates in tungsten PFCs, (b) Temperature dependence of deuterium retention at displacement damage, (c) D retention in W at elevated temperatures; (2) Permeation - (a) Gas driven permeation results for W/Mo/SiC, (b) Plasma-driven permeation test stand for TPE; and (3) Surface studies - (a) H-sensor development, (b) Adsorption of oxygen and hydrogen on beryllium surfaces.

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In this study, the PISCES-A linear plasma instrument has been used to characterize retention in several carbon fiber composites in order to better understand the factors which lead to elevated retention levels in these materials. The PISCES instrument is capable of subjecting materials to intense fluxes (up to 1022 m-2 s-1) of low energy (150 eV) D+ ions, producing conditions similar to those encountered by plasma facing components in a fusion reactor. In this investigation, three CFCs (fabricated with different manufacturing processes) are compared with the N11 composite used in the Tore Supra reactor. The specific surface areas for these materials were within the range of 0.14-0.55 m2/g. The plasma bombardment conditions were adjusted to provide doses on the order of 1025-1026 m-2 at a sample temperature of 200 °C. After removal from PISCES-A, the amount of D retained in the sample surface was determined via thermal desorption spectroscopy. The measured retention showed a strong correlation with the type of material used and the corresponding BET surface area. By using a CFC with a lower internal porosity, one could expect a reduction in retention by a factor of 5 or more. © 2008 Elsevier B.V.

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