R.A. Causey

Sandia National Laboratories, Albuquerque, New Mexico, United States

Are you R.A. Causey?

Claim your profile

Publications (138)176.32 Total impact

  • R.A. Causey · R.A. Karnesky · C. San Marchi
    [Show abstract] [Hide abstract]
    ABSTRACT: Tritium is a radioactive form of hydrogen. Because it is radioactive, its release to the environment must be minimized. Most of the materials used in fusion reactors are metals that have a relatively high permeability for tritium. The fusion community has been working on barrier materials to minimize tritium release by permeation through structural materials. Unfortunately, most barrier materials work very well during laboratory experiments, but fail to meet requirements when placed in radiation environments.This chapter presents tritium permeation characteristics of various materials used in fusion reactors, including plasma-facing, structural, and barrier materials. The parameters necessary for tritium release calculations for various regions of a fusion reactor are given.
    No preview · Article · Dec 2012
  • [Show abstract] [Hide abstract]
    ABSTRACT: Accumulation of 3He can result in a change of hydrogen isotope interactions with metal due to appearance of additional structural defects. The work is devoted to comparative study of hydrogen interactions with pure Ni and Ni containing radiogenic helium. "Tritium trick" technique was used for a build-up of radiogenic helium inside Ni samples.
    No preview · Article · Nov 2011 · Fusion Science and Technology
  • [Show abstract] [Hide abstract]
    ABSTRACT: Mechanical properties, structural changes and hydrogen interactions with stainless steel 12Cr18Ni10Ti subjected to accelerated radiogenic 3He buildup by means of "tritium trick" technique were studied. After saturation with tritium up to equilibrium concentration at a pressure 50 MPa and T=773 K the samples were rapidly cooled to room temperature and aged at this temperature up to the buildup of a predetermined 3He concentration. Kinetics of helium thermal release, hydrogen transport, trapping and accumulation in steel containing various concentration of 3He, synergistic influence of 3He and hydrogen on mechanical properties of steel containing up to 500 appm 3He and structural changes at various 3He concentrations are discussed.
    No preview · Article · Nov 2011 · Fusion Science and Technology
  • R. D. Kolasinski · D. F. Cowgill · R. A. Causey
    [Show abstract] [Hide abstract]
    ABSTRACT: 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.
    No preview · Article · Aug 2011 · Journal of Nuclear Materials
  • [Show abstract] [Hide abstract]
    ABSTRACT: The tungsten ITER divertor will be operated at temperatures above 1000K. 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.
    No preview · Article · Aug 2011 · Journal of Nuclear Materials
  • [Show abstract] [Hide abstract]
    ABSTRACT: The tritium plasma experiment (TPE) is a unique facility devoted to experiments on the behavior of deuterium/tritium in toxic (e.g., beryllium) and radioactive materials for fusion plasma-wall interaction studies. A Langmuir probe was added to the system to characterize the plasma conditions in TPE. With this new diagnostic, we found the achievable electron temperature ranged from 5.0 to 10.0 eV, the electron density varied from 5.0 × 10(16) to 2.5 × 10(18) m(-3), and the ion flux density varied between 5.0 × 10(20) to 2.5 × 10(22) m(-2) s(-1) along the centerline of the plasma. A comparison of these plasma parameters with the conditions expected for the plasma facing components (PFCs) in ITER shows that TPE is capable of achieving most (∼800 m(2) of 850 m(2) total PFCs area) of the expected ion flux density and electron density conditions.
    No preview · Article · Aug 2011 · The Review of scientific instruments
  • [Show abstract] [Hide abstract]
    ABSTRACT: This study investigates a pathway to nanoporous structures created by hydrogen and helium implantation in aluminum. Previous experiments for fusion applications have indicated that hydrogen and helium ion implantations are capable of producing bicontinuous nanoporous structures in a variety of metals. This study focuses specifically on implantations of hydrogen and helium ions at 25 keV in aluminum. The hydrogen and helium systems result in remarkably different nanostructures of aluminum at the surface. Scanning electron microscopy, focused ion beam, and transmission electron microscopy show that both implantations result in porosity that persists approximately 200 nm deep. However, hydrogen implantations tend to produce larger and more irregular voids that preferentially reside at defects. Implantations of helium at a fluence of 10{sup 18} cm{sup -2} produce much smaller porosity on the order of 10 nm that is regular and creates a bicontinuous structure in the porous region. The primary difference driving the formation of the contrasting structures is likely the relatively high mobility of hydrogen and the ability of hydrogen to form alanes that are capable of desorbing and etching Al (111) faces.
    No preview · Conference Paper · Feb 2010
  • [Show abstract] [Hide abstract]
    ABSTRACT: This study investigates a pathway to nanoporous structures created by hydrogen and helium implantation in aluminum. Previous experiments for fusion applications have indicated that hydrogen and helium ion implantations are capable of producing bicontinuous nanoporous structures in a variety of metals. This study focuses specifically on implantations of hydrogen and helium ions at 25 keV in aluminum. The hydrogen and helium systems result in remarkably different nanostructures of aluminum at the surface. Scanning electron microscopy, focused ion beam, and transmission electron microscopy show that both implantations result in porosity that persists approximately 200 nm deep. However, hydrogen implantations tend to produce larger and more irregular voids that preferentially reside at defects. Implantations of helium at a fluence of 1018 cm-2 produce much smaller porosity on the order of 10 nm that is regular and creates a bicontinuous structure in the porous region. The primary difference driving the formation of the contrasting structures is likely the relatively high mobility of hydrogen and the ability of hydrogen to form alanes that are capable of desorbing and etching Al (111) faces.
    No preview · Article · Dec 2009 · MRS Online Proceeding Library
  • [Show abstract] [Hide abstract]
    ABSTRACT: Under appropriate conditions, exposing tungsten to a high flux D plasma creates near-surface blisters and other changes in surface morphology. We have characterized the sizes of blisters formed at different temperatures (147 °C≤Tsurface≤704 °C) and performed a surface analysis to elucidate factors that influence blister formation. Tungsten targets that were exposed to low energy (70 eV) D ions at a flux of 1.1×1022 m−2 s−1 in the tritium plasma experiment (TPE) were considered. We used AES to analyze the surface for evidence of implanted impurities. Blister diameters and heights were quantified using SEM imagery and vertical scanning interferometry. Given the likelihood of D precipitation in blisters, we expect that the data obtained here could be incorporated into a computational model to better simulate the diffusion and desorption of D in W. With this in mind, we present an analysis of thermal desorption profiles showing the release of D from the surface.
    No preview · Article · Dec 2009 · Physica Scripta
  • Source
    Weifang Luo · Donald F. Cowgill · Rion A. Causey
    [Show abstract] [Hide abstract]
    ABSTRACT: Absorption isotherms at 323 K for the H−D−Pd system were measured by introducing H2 and D2 into Pd in sequence. The method using addition of isotopes to the system in sequence to investigate isotope exchange effects has not been previously reported. The equilibrium absorption pressure in the plateau region of the mixed-isotope system varies with the ratio of H/D in the solid phase. It lies between those of the single-isotope systems of H−Pd and D−Pd. Higher equilibrium pressures are associated with high D/H ratios in the solid phase. A model proposed previously (Luo, W.; Cowgill, D.; Causey, R.; Stewart, K. J. Phys. Chem., B 2008, 112, 8099) for mixed isotope hydride desorption, which correlates the equilibrium plateau pressure of the mixed H−D system with the fractions of D and H in the solid and the equilibrium plateau pressures of the single-isotope systems, is also successfully applied to absorption. When D2 is introduced into the H−Pd system in the plateau region, both the H−D exchange processes in the gas phase and net H (D) absorption take place. The former does not result in a total pressure change, but the latter creates a total pressure decrease. These reactions produce a D concentration increase in both the bulk Pd and the gaseous phase, as expected. Curiously, however, they also result in a counterintuitive small H concentration increase in bulk Pd and a decrease in gaseous H. Analogous results are obtained when the order of D2−H2 introduction is reversed. In the plateau region, isotope displacement does not take place. Once in the β-phase, isotope displacement does take place. The equilibrium isotope H−D partitions in the gas phase, H2, HD, and D2, are controlled by the equilibrium constant, KHD, and their equilibrium partitions among H and D between gas and bulk Pd are controlled by the separation factor, α.
    Full-text · Article · Nov 2009 · The Journal of Physical Chemistry C
  • Weifang Luo · Donald F Cowgill · Rion A Causey
    [Show abstract] [Hide abstract]
    ABSTRACT: A Sieverts' apparatus coupled with a residual gas analyzer (RGA) is an effective method to detect composition variations during isotopic exchange. This experimental setup provides a tool for the thermodynamic and kinetic characterization of H-D isotope exchange on Pd. The H or D concentrations in the gas and solid phases during the exchanges starting from (H(2) + Pd(x)D) and (D(2) + Pd(x)H) in beta-phase Pd were monitored over a temperature range from 173 to 298 K. The equilibrium properties, i.e., the H-D separation factors alpha and equilibrium constants K(HD), were obtained and found to be very close to those in the literature. The values of equilibrium constant reported here are the only experimental K(HD) data for H-D-Pd system. The H-D exchange rates on beta-Pd were measured for both exchange directions. A comprehensive kinetic model is proposed that correlates the exchange rate and the driving force composed of the reactant concentrations and the extent of deviation from equilibrium. The rate constants were obtained using this model for two exchange directions. The rates for the two exchange directions were found to be close to each other at 173 K, but they differ with temperature increase in such a way that the (D(2) + Pd(x)H) has a higher rate than (H(2) + Pd(x)D). The exchange activation energies obtained are 2.0 and 3.5 kJ/mol for the (H(2) + Pd(x)D) and (D(2) + Pd(x)H) directions, respectively. The difference in activation energies results from the difference in the energy states of (H(2) + Pd(x)D) and (D(2) + Pd(x)D). The calculated exchange profiles using this model agree with the experimental values reasonably well.
    No preview · Article · Oct 2009 · The Journal of Physical Chemistry B
  • [Show abstract] [Hide abstract]
    ABSTRACT: The Tritium Plasma Experiment (TPE) has been used to investigate deuterium fuel retention behavior in tungsten and molybdenum-materials utilized for plasma-facing surfaces in some existing tokamak plasma devices and under consideration for future devices. Although several studies have been performed over the past several years on these metals, many issues remain unresolved, including for example blister formation mechanisms and correlation to surface conditions. In this study, we expose several metal samples to deuterium ion fluences up to 1026 ions/m2 and measure retention behavior with thermal desorption spectroscopy. Fractional retention of up to 2.0×10−5 is found for W at 600K, and Mo similarly retains deuterium at a fraction of 1.5×10−5 at 600K. Blistering was found for W samples exposed at temperatures above 453K, whereas blistering was not observed for Mo samples at any experiment temperature.
    No preview · Article · Jun 2009 · Journal of Nuclear Materials
  • [Show abstract] [Hide abstract]
    ABSTRACT: 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.
    No preview · Article · Jun 2009 · Fusion Engineering and Design
  • Source
    [Show abstract] [Hide abstract]
    ABSTRACT: Plasma wall interaction (PWI) is important for the material choice in ITER and for the plasma scenarios compatible with material constraints. In this paper, different aspects of the PWI are assessed in their importance for the initial wall materials choice: CFC for the strike point tiles, W in the divertor and baffle and Be on the first wall. Further material options are addressed for comparison, such as W divertor/Be first wall and all-W or all-C. One main parameter in this evaluation is the particle flux to the main vessel wall. One detailed plasma scenario exists for a Q = 10 ITER discharge [G. Federici et al., J. Nucl. Mater. 290–293 (2001) 260] which was taken as the basis of further erosion and tritium retention evaluations. As the assessment of steady state wall fluxes from a scaling of present fusion devices indicates that global wall fluxes may be a factor of 4 ± 3 higher, this margin has been adopted as uncertainty of the scaling. With these wall and divertor fluxes, important PWI processes such as erosion and tritium accumulation have been evaluated: It was found that the steady state erosion is no problem for the lifetime of plasma-facing divertor components. Be wall erosion may pose a problem in case of a concentration of the wall fluxes to small wall areas. ELM erosion may drastically limit the PFC lifetime if ELMs are not mitigated to energies below 0.5 MJ. Dust generation is still a process which requires more attention. Conversion from gross or net erosion to dust and the assessment of dust on hot surfaces need to be investigated. For low-Z materials the build-up of the tritium inventory is dominated by co-deposition with eroded wall atoms. For W, where erosion and tritium co-deposition are small, the implantation, diffusion and bulk trapping constitute the dominant retention processes. First extrapolations with models based on laboratory data show small contributions to the inventory. For later ITER phases and the extrapolation to DEMO additional tritium trapping sites due to neutron-irradiation damage need to be taken into account. Finally, the expected values for erosion and tritium retention are compared to the ITER administrative limits for the lifetime, dust and tritium inventory.
    Full-text · Article · Jun 2009 · Journal of Nuclear Materials
  • [Show abstract] [Hide abstract]
    ABSTRACT: Recent work on hydrogen isotope retention in tungsten has shown a substantial fraction of the retained hydrogen to be in the form of molecules. It can be expected that hydrogen permeating through a material such as tungsten, that has a very low solubility for hydrogen, would come out of solution and combine into molecules at voids located throughout the bulk. The purpose of this report is to determine the type of voids responsible for the molecular retention. High purity tungsten provided by Plansee Aktiengesellschaft was first polished, annealed at 1273K in vacuum for one hour, and then exposed to high fluxes and high fluences of deuterium in the PISCES facility. High resolution Transmission Electron Microscopy was then used to examine the samples for voids. The results of these experiments were used to interpret the expected behavior of tungsten to be used as the divertor of the ITER fusion device.
    No preview · Article · Jun 2009 · Journal of Nuclear Materials
  • [Show abstract] [Hide abstract]
    ABSTRACT: Management of tritium inventory remains one of the grand challenges in the development of fusion energy, and the choice of plasma-facing materials is a key factor for in-vessel tritium retention. The Atomic and Molecular Data Unit of the International Atomic Energy Agency organized a Coordinated Research Project (CRP) on the overall topic of tritium inventory infusion reactors during the period 2001-2006. This dealt with hydrogenic retention in ITER's plasma-facing materials - Be, C, and W - and in compounds (mixed materials) of these elements as well as tritium removal techniques. The results of the CRP are summarized in this paper together with recommendations for ITER. Basic parameters of diffusivity, solubility, and trapping in Be, C, and W are reviewed. For Be, the development of open porosity can account for transient hydrogenic pumping, but long-term retention will be dominated by codeposition. Codeposition is also the dominant retention mechanism for carbon and remains a serious concern for both Be- and C-containing layers. Hydrogenic trapping in unirradiated tungsten is low but will increase with ion and neutron damage. Mixed materials will be formed in a tokamak, and these can also retain significant amounts of hydrogen isotopes. Oxidative and photon-based techniques for detritiation of plasma-facing components are described.
    No preview · Article · Nov 2008 · Fusion Science and Technology
  • [Show abstract] [Hide abstract]
    ABSTRACT: Interactions between the plasma and the vessel walls constitute a major engineering problem for next step fusion devices, such as ITER, determining the choice of the plasma-facing materials. A prominent issue in this choice is the tritium inventory build-up in the vessel, which must be limited for safety reasons. The initial material selection, i.e. beryllium (Be) on the main vessel walls, tungsten (W) on the divertor upper baffle and dome, and carbon fibre composite around the strike points on the divertor plates, results both from the attempt to reduce the tritium inventory and to optimize the lifetime of the plasma-facing components.In the framework of the EU Task Force on Plasma–Wall Interaction (PWI TF), the many physics aspects governing the tritium inventory are brought together. Together with supporting information from international experts represented by the ITPA SOL/DIV section, this paper describes the present status of knowledge of the in-vessel tritium inventory build-up. Firstly, the main results from present fusion devices in this field are briefly reviewed. Then, the processes involved are discussed: implantation, trapping and diffusion in plasma-facing materials are considered as well as surface erosion and co-deposition of tritium with eroded material. The intermixing of the different materials and its influence on hydrogen retention and co-deposition is a major source of uncertainty on present estimates and is also addressed.Based on the previous considerations, estimates for the tritium inventory build-up are given for the initial choice of ITER materials, as well as for alternative options. Present estimates indicate a build-up of the tritium inventory to the administrative limit within a few hundred nominal full power D : T discharges, co-deposition with carbon being the dominant process. Therefore, tritium removal methods are also an active area of research within the EU PWI TF, and are discussed. An integrated operational scheme to slow the rate of tritium accumulation is presented, which includes plasma operation as well as conditioning procedures.
    No preview · Article · Aug 2008 · Plasma Physics and Controlled Fusion
  • [Show abstract] [Hide abstract]
    ABSTRACT: The tritium trick technique was used to build-up radiogenic helium inside stainless steel 12Cr18Ni10Ti (SS). A great quantity of defects with a mean diameter of 20 nm, most probably platelet-like bubbles with 3He atoms, was observed in 3He-containing samples. The mean density of these bubbles in SS samples containing ∼75 appm of 3He is estimated to be 6·1020 m-3. Much larger helium bubbles were observed in SS after annealing the samples at T ≥1170 K. Thermal release of radiogenic helium occurs at T>1500 K The presence of 3He in structural materials causes the formation of an additional state for hydrogen sorption.
    No preview · Article · Aug 2008 · Fusion Science and Technology
  • [Show abstract] [Hide abstract]
    ABSTRACT: Samples of stainless steel 12Cr18Ni10Ti with radiogenic helium were subjected to mechanical tests with a constant extension rate. The presence of 3He does not markedly affect the strength characteristic, but significantly decreases plasticity of steel. The presence of hydrogen enhances the embrittlement of steel, containing 3He. The diffusion coefficient of hydrogen does not change significantly in the presence of helium, but the traps for hydrogen, which occur due to the presence of helium, delay the kinetics of a steady state flux onset at helium concentration of 50 appm.
    No preview · Article · Aug 2008 · Fusion Science and Technology
  • Weifang Luo · Don Cowgill · Rion Causey · Ken Stewart
    [Show abstract] [Hide abstract]
    ABSTRACT: A Sieverts' apparatus coupled with a residual gas analysis is used to measure the concentration variations of hydrogen isotopes in the gas and solid phases during exchange and isothermal decomposition of mixed hydrides. beta-phase palladium hydrides with known ratios of H:D, Pd(H x D 1- x ) y (0 < x < 1, y > 0.6), are prepared by H 2 with PdD y or D 2 with PdH y exchange, and their desorption isotherms are reported here at 323 K. A higher equilibrium pressure in isothermal desorption of mixed hydrides is associated with a higher ratio of D/H in the initial mixed hydrides in beta-phase. The composition of the gas desorbed from a mixed hydride varies; i.e., the ratio of D/H in gas decreases with the sum of (H + D) in Pd. The values of the separation factor alpha during desorption at 323 K and during H-D exchange at 248 K are discussed and compared with those in the literature. Desorption isotherms of mixed isotope hydrides are between those of the single isotope hydrides of H-Pd and D-Pd, however, plateaus slope more than those of pure isotope hydrides. The origin of the plateau sloping in the mixed hydrides can be attributed to the compositional variations during desorption, i.e., the equilibrium pressure is greater when D/H ratio in solid is greater. A simple model is proposed in this study that agrees well with experimental results.
    No preview · Article · Jul 2008 · The Journal of Physical Chemistry B

Publication Stats

3k Citations
176.32 Total Impact Points

Institutions

  • 1984-2011
    • Sandia National Laboratories
      • Semiconductor Material and Device Sciences Department
      Albuquerque, New Mexico, United States
  • 2002
    • Johns Hopkins University
      • Department of Physics and Astronomy
      Baltimore, Maryland, United States
  • 2001
    • Los Alamos National Laboratory
      Лос-Аламос, California, United States
  • 2000
    • New Mexico Institute of Mining and Technology
      Сокорро, New Mexico, United States
  • 1989
    • Princeton University
      • Princeton Plasma Physics Laboratory
      Princeton, NJ, United States
  • 1987
    • University of California, Los Angeles
      • Department of Mechanical and Aerospace Engineering
      Los Ángeles, California, United States