R. A. Causey

Sandia National Laboratories, Albuquerque, New Mexico, United States

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Publications (128)140.62 Total impact

  • [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.
    The Review of scientific instruments 08/2011; 82(8):083503. · 1.52 Impact Factor
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    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.
    Journal of Nuclear Materials 08/2011; 415(1). · 2.02 Impact Factor
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    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.
    Journal of Nuclear Materials 01/2011; 415(1). · 2.02 Impact Factor
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    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.
    Physica Scripta 12/2009; 2009(T138):014042. · 1.03 Impact Factor
  • Weifang Luo, Donald F. Cowgill, Rion A. Causey
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    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.
    The Journal of Physical Chemistry C 11/2009; · 4.84 Impact Factor
  • Weifang Luo, Donald F Cowgill, Rion A Causey
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    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.
    The Journal of Physical Chemistry B 10/2009; 113(39):12978-87. · 3.61 Impact Factor
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    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.
    Journal of Nuclear Materials 06/2009; 390:709-712. · 2.02 Impact Factor
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    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.
    Journal of Nuclear Materials. 01/2009;
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    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.
    Journal of Nuclear Materials - J NUCL MATER. 01/2009; 390:717-720.
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    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.
    Fusion Engineering and Design. 01/2009; 84:1068-1071.
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    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.
    Plasma Physics and Controlled Fusion 08/2008; 50(10):103001. · 2.37 Impact Factor
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    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.
    The Journal of Physical Chemistry B 07/2008; 112(27):8099-105. · 3.61 Impact Factor
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    ABSTRACT: Depth profiles of deuterium trapped in tungsten exposed to a low-energy (≈200 eV/D) and high deuterium ion flux (about 1 × 1021 D/m2 s) in clean (We use the term ‘clean’ in quotation marks having in mind the impossibility to obtain absolutely clean plasma. In our case the conception ‘clean’ D plasma means the plasma without intentionally introduced carbon impurities.) and carbon-seeded D plasmas at an ion fluence of about 2 × 1024 D/m2 and various temperatures have been measured up to a depth of 7 μm using the D(3He, p)4He nuclear reaction at a 3He energy varied from 0.69 to 4.0 MeV. The deuterium retention in single-crystalline and polycrystalline W increases with the exposure temperature, reaching its maximum value at about 500 K (for ‘clean’ plasma) or about 600 K (for carbon-seeded plasma), and then decreases as the temperature grows further. It is assumed that tungsten carbide formed on the W surface under exposure to the carbon-seeded D plasmas serves as a barrier layer for diffusion and prevents the outward transport of deuterium, thus increasing the D retention in the bulk of tungsten.
    Journal of Nuclear Materials 04/2008; · 2.02 Impact Factor
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    ABSTRACT: The Tritium Plasma Experiment (TPE) has been recently relocated from Los Alamos National Laboratory (LANL) to Safety and Tritium Applied Research (STAR) facility in Idaho National Laboratory (INL). The application of a Langmuir probe system, newly designed target holder, and thermal desorption spectroscopy (TDS) system were successfully carried out, and the initial results from Langmuir probe measurements in deuterium plasma and TDS measurements of deuterium retention in tungsten are discussed. TPE is now ready to provide data to the fusion community on the interaction of tritium plasma with plasma facing components, and the future research plan is discussed.
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    ABSTRACT: We present some results of studying the influence of high-pressure hydrogen (80 MPa), radiogenic 3He (with concentrations up to 130 appm), and their joint action on the mechanical properties and structure of 12Kh18N10T steel in the temperature range from 293 to 873 K. We describe the procedure of tests of specimens containing 3He. It has been established that the joint action of hydrogen and 3He affects slightly the ultimate strength of the specimens. Saturation of steel with radiogenic 3He by the method of “tritium trick” increases its yield strength. Hardening of the steel caused by helium increases with temperature and 3He concentration and, at 873 K, is accompanied by substantial embrittlement. We also present results of the fractographic analysis of specimens tested under different conditions.
    Materials Science 09/2007; 43(5). · 0.21 Impact Factor
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    ABSTRACT: 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 1μm. 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.
    Journal of Nuclear Materials 08/2007; 367:1512-1515. · 2.02 Impact Factor
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    ABSTRACT: Collaborative Be–W interaction experiments conducted at UC-San Diego PISCES Laboratory, and Sandia National Laboratories, Livermore, CA (SNL/CA), are reported. In the divertor-plasma simulator PISCES–B, W targets are exposed to Be seeded D2 plasma in the temperature range 1070–1320K. All reveal the formation of surface Be–W alloying. The alloy reaction rate is found to increase with surface temperature in the range 1023–1123K in SNL vacuum-deposition phase formation experiments. In both sets of experiments the efficiency of surface alloying is found to depend on the availability of surface deposited Be. This availability is reduced by evaporation at high temperature, and also by plasma re-erosion in the case of PISCES–B targets. Surface analysis of targets using Auger electron spectroscopy (AES), wavelength dispersive X-ray spectroscopy (WDS), and X-ray diffraction (XRD) reveals Be12W as the dominant alloy composition where Be surface availability is optimal.
    Journal of Nuclear Materials 06/2007; 363:1179-1183. · 2.02 Impact Factor
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    ABSTRACT: Tritium retention measurements have been performed for two carbon composites both before and after neutron irradiation. These two composites were FMI-222 manufactured by Fiber Materials Incorporated and MKC-1PH by Mitsubishi Kasei. The neutron irradiations were performed in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory at 473 K and spanned radiation levels from unirradiated up to 1.0 dpa. The high energy traps in the composites were saturated in tritium by heating the composites to 1673 K with a pure tritium gas pressure of 13.2 Pa. The samples were then outgassed in a separate system at a temperature of 1963 K where tritium analysis was performed using an ionization chamber and liquid scintillation counting. For all irradiation conditions, the retained tritium was less than that measured in earlier studies for different types of graphites. The results suggest carbon composites should be preferred over graphites for use in fusion reactors where both tritium and neutron irradiation exist.
    Physica Scripta 12/2006; 1996(T64):32. · 1.03 Impact Factor
  • B.M. Oliver, Y. Dai, R.A. Causey
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    ABSTRACT: Three irradiations have been performed in the Swiss Spallation Neutron Source (SINQ) to establish a materials database for mixed proton and neutron fluxes for future spallation neutron and other accelerator sources. Samples of 316LN, F82H, AlMg3, and Zircaloy-2 from STIP-II have been analyzed for their total helium and hydrogen contents and their release characteristics with temperature. Helium and hydrogen release measurements showed considerable levels of deuterium and tritium species which generally mirrored those of hydrogen. Hydrogen release occurred from about 300 °C for the AlMg3 to about 800 °C for the Zircaloy-2. For the Zircaloy-2 and the steels, helium release began to occur at between 1100 and 1200 °C, which is consistent with previous measurements on irradiated steels. Modeling of the hydrogen release data for the 316 and F82H suggests two traps of differing energy dependent on the irradiation dose and temperature. The higher energy traps are probably voids created from vacancy coalescence.
    Journal of Nuclear Materials 09/2006; 356(s 1–3):148–156. · 2.02 Impact Factor
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    ABSTRACT: For the Accelerator Production of Tritium Project (APT), spallation neutrons will be moderated and then absorbed in 3He gas to produce tritium. The spallation neutrons will be generated by the interaction of high energy (~1 GeV) protons with solid tungsten rods or cylinders. A byproduct of the spallation reactions is large amounts of helium and hydrogen gas generated in the rods and other structural materials. The release kinetics of these gases during various proposed off-normal scenarios involving loss of coolant and afterheat-induced rises in temperature is of particular interest to the APT Project. In addition, however, this data is of interest for fusion reactors where tungsten used in a tokamak divertor will also be exposed to neutrons. In this case, the generated protium will be accompanied by deuterium and tritium diffusing in from the plasma-facing surface. Tungsten rods irradiated with 800 MeV protons in the Los Alamos Neutron Science Center (LANCE) to high exposures have been sectioned to produce small specimens suitable for measurement of both hydrogen and helium. Hydrogen evolution was measured by subjecting the specimens to a simulated temperature ramp from ~200 to ~1200°C, similar to that expected due to a loss of coolant and subsequent afterheat. The release measurements were conducted using mass spectrometric techniques. Four release peaks at temperatures of approximately 550, 850, 1100 and 1200°C were observed, initially suggesting a number of trapping sites with different binding energies. Subsequent analysis, however, showed that the observed peaks were artifacts of the temperature heating profile, and that the release curve could be duplicated using a single trap energy of 1.4 eV.
    Physica Scripta 08/2006; 2001(T94):137. · 1.03 Impact Factor

Publication Stats

1k Citations
140.62 Total Impact Points


  • 1984–2011
    • Sandia National Laboratories
      Albuquerque, New Mexico, United States
  • 2005
    • University of California, San Diego
      • Department of Mechanical and Aerospace Engineering (MAE)
      San Diego, CA, United States
  • 1989
    • Princeton University
      • Princeton Plasma Physics Laboratory
      Princeton, NJ, United States