The helium-related characteristics of aged palladium tritide, including tritium loss, the formation of He bubbles from tritium decay-induced He atoms, and He thermal desorption behavior are computed using a continuum model of bubble evolution that incorporates the log-normal bubble spacing distribution deduced from 3 He nuclear magnetic resonance (NMR) measurements. This spacing distribution produces significant differences between mean (expectation) values of the bubble size and pressure and the average (median) values of these aging characteristics obtained by a previous calculation. The new calculations find much of the He is retained in a large quantity of smaller bubbles with higher bubble pressures, improving the overall stability of the bubbles and extending the age of He retention. By contrast, the integrated tritide material characteristics of swelling and tritium pressure are found relatively insensitive to details of the bubble spacing distribution. Inter-bubble fracture is predicted to begin with small, closely-spaced bubbles and progress to include larger bubbles near the onset of rapid He release. The critical concentration for this onset increases almost linearly with material tensile strength and decreases with shear modulus. Release of He from the aging solid is modeled as a growing network of fractured inter-bubble ligaments. For small particle or small grain material, the physical size of the fracture cluster, relative to the material dimensions, becomes important. In contrast with normal aging, inter-bubble fracture occurring during thermal desorption appears to begin with large, widely-spaced bubbles, indicating elevated temperature techniques may be of limited use for evaluating development of the bubble fracture network.
Tritium for the U.S. nuclear weapon stockpile is produced in tritium producing burnable absorber rods (TPBARs) inserted into Tennessee Valley Authoritys (TVA) light-water nuclear reactors. The rods are stainless steel tubes with a permeation barrier coating and internal components that generate and contain the tritium. The TPBAR incorporates a Ni-plated Zircoloy getter tube to capture tritium and prevent it from reaching the rod cladding and permeating into the environment. Under the conventional view of getter behavior, the tritium pressure outside the getter tube is expected to be limited to the equilibrium vapor pressure of Zr hydride at the temperature of the rod as long as the total hydrogen concentration remains below the capacity of the hydride. Since the tritium pressure is higher within the rod core, this behavior relies on the thin getters ability to hold off a differential tritium pressure. The effective tritium pressure on the cladding can also be enhanced by isotope exchange. Hydrogen ingress through the cladding from the reactor coolant creates a hydrogen pressure on the outer surface of the getter that can exchange with tritium, allowing the tritium partial pressure to increase toward this hydrogen gettering pressure. The goal of this work was to use laboratory-scale experiments to examine these mechanisms and create a model of getter behavior that describes tritium transport within the TPBAR. A third mechanism wherein the concentration at the outer surface of the getter is increased by the temperature gradient within the getter tube wall (the Soret effect) is not experimentally tested but is captured in the model. While not conclusively demonstrated by the experiments due to low pressure, high temperature, and small gap volume conditions, the model shows that when combined, the three mechanisms can explain both the magnitude and time dependence of the tritium release observed for reactor fuel assemblies with TPBARs. The model also shows how various modifications of the TPBAR design can reduce this tritium release into the environment.
We examine how deuterium becomes trapped in plasma-exposed tungsten and forms near-surface platelet-shaped precipitates. How these bubbles nucleate and grow, as well as the amount of deuterium trapped within, is crucial for interpreting the experimental database. Here, we use a combined experimental/theoretical approach to provide further insight into the underlying physics. With the Tritium Plasma Experiment, we exposed a series of ITER-gradetungsten samples to high flux D plasmas (up to 1.5 × 1022 m-2 s-1) at temperatures ranging between 103 and 554 °C. Retention of deuterium trapped in the bulk, assessed through thermal desorption spectrometry, reached a maximum at 230 °C and diminished rapidly thereafter for T > 300 °C. Post-mortem examination of the surfaces revealed non-uniform growth of bubbles ranging in diameter between 1 and 10 μm over the surface with a clear correlation with grain boundaries. Electron back-scattering diffraction maps over a large area of the surface confirmed this dependence; grains containing bubbles were aligned with a preferred slip vector along the <111> directions. Focused ion beam profiles suggest that these bubbles nucleated as platelets at depths of 200 nm–1 μm beneath the surface and grew as a result of expansion of sub-surface cracks. Furthermore, to estimate the amount of deuterium trapped in these defects relative to other sites within the material, we applied a continuum-scale treatment of hydrogen isotope precipitation. Additionally, we propose a straightforward model of near-surface platelet expansion that reproduces bubble sizes consistent with our measurements. For the tungsten microstructure considered here, we find that bubbles would only weakly affect migration of D into the material, perhaps explaining why deep trapping was observed in prior studies with plasma-exposed neutron-irradiated specimens. We foresee no insurmountable issues that would prevent the theoretical framework developed here from being extended to a broader range of systems where precipitation of insoluble gases in ion beam or plasma-exposed metals is of interest.
Hydrogen isotope gas exchange within palladium powders is examined using a batch-type reactor coupled to a residual gas analyzer (RGA). Exchange rates in both directions (H2 + PdD and D2 + PdH) are measured in the temperature range 178-323 K for the samples with different particle sizes. The results show this batch-type exchange is closely approximated as a first-order kinetic process with a rate directly proportional to the surface area of the powder particles. An exchange rate constant of 1.40 ± 0.24 μmol H2/atm cm2 s is found for H2 + PdD at 298 K, 1.4 times higher than that for D2 + PdH, with an activation energy of 25.0 ± 3.2 kJ/mol H for both exchange directions. A comparison of exchange measurement techniques shows these coefficients, and the fundamental exchange probabilities are in good agreement with those obtained by NMR and flow techniques.
In this study, the authors developed an approach for accurately quantifying the helium content in a gas mixture also containing hydrogen and methane using commercially available getters. The authors performed a systematic study to examine how both H2 and CH4 can be removed simultaneously from the mixture using two SAES St 172® getters operating at different temperatures. The remaining He within the gas mixture can then be measured directly using a capacitance manometer. The optimum combination involved operating one getter at 650 °C to decompose the methane, and the second at 110 °C to remove the hydrogen. This approach eliminated the need to reactivate the getters between measurements, thereby enabling multiple measurements to be made within a short time interval, with accuracy better than 1%. The authors anticipate that such an approach will be particularly useful for quantifying the He-3 in mixtures that include tritium, tritiated methane, and helium-3. The presence of tritiated methane, generated by tritium activity, often complicates such measurements.
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.
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.
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.
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}).
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.
A continuum-scale, evolutionary model of bubble nucleation, growth and He release for aging metal tritides is described which accounts for major features of the tritide database. Bubble nucleation, modeled as self-trapping of interstitially diffusing He atoms, occurs during the first few days following tritium introduction into the metal. Bubble growth by dislocation loop punching yields good agreement between He atomic volumes and bubble pressures determined from bulk swelling and 3He NMR data. The bubble spacing distribution determined from NMR is shown to remain fixed with age, justifying the separation of nucleation and growth phases and providing a sensitive test of the growth formulation. Late in life, bubble interactions are proposed to produce cooperative stress effects, which lower the bubble pressure. Helium generated near surfaces and surface-connected porosity accounts for the low-level early helium release. Use of an average ligament stress criterion predicts an onset of inter-bubble fracture in good agreement with the He/Metal ratio observed for rapid He release. From the model, it is concluded that He retention can be controlled through control of bubble nucleation.
A continuum-scale, evolutionary model of bubble nucleation, growth and He release for aging metal tritides is described which accounts for major features of the tritide database. Bubble nucleation, modeled as self-trapping of interstitially diffusing He atoms, occurs during the first few days following tritium introduction into the metal. Bubble growth by dislocation loop punching yields good agreement between He atomic volumes and bubble pressures determined from bulk swelling and 3He NMR data. The bubble spacing distribution determined from NMR is shown to remain fixed with age, justifying the separation of nucleation and growth phases and providing a sensitive test of the growth formulation. Late in life, bubble interactions are proposed to produce cooperative stress effects, which lower the bubble pressure. Helium generated near surfaces and surface-connected porosity accounts for the low-level early helium release. Use of an average ligament stress criterion predicts an onset of inter-bubble fracture in good agreement with the He/Metal ratio observed for rapid He release. From the model, it is concluded that He retention can be controlled through control of bubble nucleation.
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.
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.