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.

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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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Nano-structured palladium is examined as a tritium storage material with the potential to release beta-decay-generated helium at the generation rate, thereby mitigating the aging effects produced by enlarging He bubbles. Helium retention in proposed structures is modeled by adapting the Sandia Bubble Evolution model to nano-dimensional material. The model shows that even with ligament dimensions of 6-12 nm, elevated temperatures will be required for low He retention. Two nanomaterial synthesis pathways were explored: de-alloying and surfactant templating. For de-alloying, PdAg alloys with piranha etchants appeared likely to generate the desired morphology with some additional development effort. Nano-structured 50 nm Pd particles with 2-3 nm pores were successfully produced by surfactant templating using PdCl salts and an oligo(ethylene oxide) hexadecyl ether surfactant. Tests were performed on this material to investigate processes for removing residual pore fluids and to examine the thermal stability of pores. A tritium manifold was fabricated to measure the early He release behavior of this and Pd black material and is installed in the Tritium Science Station glove box at LLNL. Pressure-composition isotherms and particle sizes of a commercial Pd black were measured.

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Erbium is used as a storage medium for tritium. Microstructural study of helium bubble generation from tritium decay in erbium tritide can provide an unusual example of bubble development with negligible radiation damage. Aged erbium tritide film specimens were found to contain five distinctly different microstructural features. The general structure was of large columnar grains of ErT2. But on a fine scale, precipitates believed to be erbium oxy-tritides and helium bubbles could be identified. The precipitate size was in the range of ∼10 nm and the bubbles were of an unusual planar shape on {111} planes with an invariant thickness of ∼1 nm and a diameter on the order of 10 nm. Also, an outer layer containing no fine precipitate structure and only a few helium bubbles were present on the films. This layer is best described as a denuded zone which probably grew during aging in air. Finally, large embedded Er2O3 particles were found at low density and nonuniformly distributed, but sometimes extending through the thickness of the film. A failure mechanism allowing the helium to escape is suggested by observed cracking between bubbles closer to end of life. Copyright © 2007 by ASTM International.

In this study we examine the combination of a He-O glow discharge with heating as a possible technique to remove deuterium from TFTR tiles. Samples were cut from a relatively large area containing a uniform codeposited layer of deuterium and carbon. Auger/SEM was used to generate micrographs of each of the samples. The samples were also examined using Rutherford backscattering to determine the near surface composition. Individual samples were then exposed to a He-O glow discharge while being heated. After the exposure, the samples were returned for Auger/SEM and RBS of the same areas examined prior to the exposure. Comparing the samples before and after exposure revealed that the amount of the codeposited layer removed was significantly less than 1 μm. Removal rates this low would suggest that He-O glow discharge with heating is insufficient to remove the thick layers predicted for ITER in a timely fashion.

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A model is presented for the linking of helium bubbles growing in aging metal tritides. Stresses created by neighboring bubbles are found to produce bubble growth toward coalescence. This process is interrupted by the fracture of ligaments between bubble arrays. The condition for ligament fracture percolates through the material to reach external surfaces, leading to material micro-cracking and the release of helium within the linked-bubble cluster. A comparison of pure coalescence and pure fracture mechanisms shows the critical HeM concentration for bubble linkage is not strongly dependent on details of the linkage process. The combined stress-directed growth and fracture process produces predictions for the onset of rapid He release and the He emission rate. Transition to this rapid release state is determined from the physical size of the linked-bubble clusters, which is calculated from dimensional invariants in classical percolation theory. The result is a transition that depends on material dimensions. The onset of bubble linkage and rapid He release are found to be quite sensitive to the bubble spacing distribution, which is log-normal for bubbles nucleated by self-trapping.

The oxidation of zirconium alloys is one of the most studied processes in the nuclear industry. The purpose of this report is to provide in a concise form a review of the oxidation process of zirconium alloys in the moderate temperature regime. In the initial ''pre-transition'' phase, the surface oxide is dense and protective. After the oxide layer has grown to a thickness of 2 to 3 {micro}m's, the oxidation process enters the ''post-transition'' phase where the density of the layer decreases and becomes less protective. A compilation of relevant data suggests a single expression can be used to describe the post-transition oxidation rate of most zirconium alloys during exposure to oxygen, air, or water vapor. That expression is: Oxidation Rate = 13.9 g/(cm{sup 2}-s-atm{sup -1/6}) exp(-1.47 eV/kT) x P{sup 1/6} (atm{sup 1/6}).

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Sandia National Laboratories in California initiated an experimental program to determine whether tritium retention in the tube walls and permeation through the tubes into the surrounding coolant water would be a problem for the Accelerator Production of Tritium (APT), and to find ways to mitigate the problem, if it existed. Significant holdup in the tube walls would limit the ability of APT to meet its production goals, and high levels of permeation would require a costly cleanup system for the cooling water. To simulate tritium implantation, a 200 keV accelerator was used to implant deuterium into Al 6061-T and SS3 16L samples at temperatures and particle fluxes appropriate for APT, for times varying between one week and five months. The implanted samples were characterized to determine the deuterium retention and Permeation. During the implantation, the D(d,p)T nuclear reaction was used to monitor the build-up of deuterium in the implant region of the samples. These experiments increased in sophistication, from mono-energetic deuteron implants to multi-energetic deuteron and proton implants, to more accurately reproduce the conditions expected in APT. Micron-thick copper, nickel, and anodized aluminum coatings were applied to the front surface of the samples (inside of the APT walls) in an attempt to lower retention and permeation. The reduction in both retention and permeation produced by the nickel coatings, and the ability to apply them to the inside of the APT tubes, indicate that both nickel-coated Al 6061-T6 and nickel-coated SS3 16L tubes would be effective for use in APT. The results of this work were submitted to the Accelerator Production of Tritium project in document number TPO-E29-Z-TNS-X-00050, APT-MP-01-17.

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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, sodium and beryllium 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 3D 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. © 2004 Elsevier B.V. All rights reserved.

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Accurate modeling of nucleation, growth and clustering of helium bubbles within metal tritide alloys is of high scientific and technological importance. Of interest is the ability to predict both the distribution of these bubbles and the manner in which these bubbles interact at a critical concentration of helium-to-metal atoms to produce an accelerated release of helium gas. One technique that has been used in the past to model these materials, and again revisited in this research, is percolation theory. Previous efforts have used classical percolation theory to qualitatively and quantitatively model the behavior of interstitial helium atoms in a metal tritide lattice; however, higher fidelity models are needed to predict the distribution of helium bubbles and include features that capture the underlying physical mechanisms present in these materials. In this work, we enhance classical percolation theory by developing the dynamic point-source percolation model. This model alters the traditionally binary character of site occupation probabilities by enabling them to vary depending on proximity to existing occupied sites, i.e. nucleated bubbles. This revised model produces characteristics for one and two dimensional systems that are extremely comparable with measurements from three dimensional physical samples. Future directions for continued development of the dynamic model are also outlined.

A continuum-scale, evolutionary model of helium (He) nano-bubble nucleation, growth and He release for aging bulk metal tritides is presented which accounts for major features of the experimental database. Bubble nucleation, modeled as self-trapping of interstitially diffusing He atoms, is found to occur during the first few days following tritium introduction into the metal and is sensitive to the He diffusivity and pairing energy. An effective helium diffusivity of 0.3 x 10{sup -16} cm{sup 2}/s at 300 K is required to generate the average bubble density of 5x 1017 bubbles/cm3 observed by transmission electron microscopy (TEM). Early bubble growth by dislocation loop punching with a l/radius bubble pressure dependence produces good agreement with He atomic volumes and bubble pressures determined from swelling data, nuclear magnetic resonance (NMR) measurements, and hydride pressure-composition-temperature (PCT) shifts. The model predicts that later in life neighboring bubble interactions may first lower the loop punching pressure through cooperative stress effects, then raise the pressure by partial blocking of loops. It also accounts for the shape of the bubble spacing distribution obtained from NMR data. This distribution is found to remain fixed with age, justifying the separation of nucleation and growth phases, providing a sensitive test of the growth formulation, and indicating that further significant bubble nucleation does not occur throughout life. Helium generated within the escape depth of surfaces and surface-connected porosity produces the low-level early helium release. Accelerated or rapid release is modeled as inter-bubble fracture using an average ligament stress criterion. Good agreement is found between the predicted onset of fracture and the observed He-metal ratio (HeM) for rapid He release from bulk palladium tritide. An examination of how inter-bubble fracture varies over the bubble spacing distribution shows that the critical Hem will be lower for thin films or small particle material. It is concluded that control of He retention can be accomplished through control of bubble nucleation.