W.R. Wampler

Ion-beam studies of hydrogen-metal interactions

Abstract Experiments utilizing ion implantation and ion-beam analysis have provided a large body of new quantitative information on hydrogen interactions within metal matrices and at metal surfaces. Investigated matrix interactions include trapping by vacancies, vacancy-solute complexes, He bubbles, and oxide precipitates, together with such phase-change reactions as hydride formation, hydrogen-bubble precipitation, and hydrogen reduction of precipitated oxides. Extracted information encompasses mechanisms, binding enthalpies, microstructures, and local atomic configurations. In the area of surfaces, ion-channeling analysis has yielded the lattice positions of chemisorbed hydrogen on a variety of metals. Additionally, ion-implantation experiments have quantified surface-limited uptake and release for bare, chemisorbed, and oxidized metal surfaces. Theoretical studies coupled with the ion-beam experiments have produced advances in four areas of hydrogen behavior. First, effective-medium theory has been shown to yield a quantitative description of hydrogen energies and positions at matrix defect traps and surface chemisorption sites. Second, a new, analytical treatment of trapping kinetics gives quantitative trapping rates for arbitrary trap volume fraction and arbitrary spatial correlation between traps. Third, new theoretical studies of surface barriers have clarified the dominant processes of release for bare, chemisorbed, and oxidized metal surfaces. Finally, general transport formalisms have been developed to predict overall hydrogen behavior in terms of individual matrix and surface processes.

Tritium retention in tungsten exposed to intense fluxes of 100 eV tritons

Abstract Tungsten is a candidate material for the International Thermonuclear Experimental Reactor (ITER) as well as other future magnetic fusion energy devices. Tungsten is well suited for certain fusion applications in that it has a high threshold for sputtering as well as a very high melting point. As with all materials to be used on the inside of a tokamak or similar device, there is a need to know the behavior of hydrogen isotopes embedded in the material. With this need in mind, the Tritium Plasma Experiment (TPE) has been used to examine the retention of tritium in tungsten exposed to very high fluxes of 100 eV tritons. Both tungsten and tungsten containing 1% lanthanum oxide were used in these experiments. Measurements were performed over the temperature range of 423–973 K. After exposure to the tritium plasma, the samples were transferred to an outgassing system containing an ionization chamber for detection of the released tritium. The samples were outgassed using linear ramps from room temperature up to 1473 K. Unlike most other materials exposed to energetic tritium, the tritium retention in tungsten reaches a maximum at intermediate temperatures with low retention at both high and low temperatures. For the very high triton fluences used (>10 25 T/m 2 ), the fractional retention of the tritium was below 0.02% of the incident particles. This report presents not only the results of the tritium retention, but also includes the modeling of the results and the implication for ITER and other future fusion devices where tungsten is used.

The effect of lithium surface coatings on plasma performance in the National Spherical Torus Experiment

National Spherical Torus Experiment [which M. Ono et al., Nucl. Fusion 40, 557 (2000)] high-power divertor plasma experiments have shown, for the first time, that benefits from lithium coatings applied to plasma facing components found previously in limited plasmas can occur also in high-power diverted configurations. Lithium coatings were applied with pellets injected into helium discharges, and also with an oven that directed a collimated stream of lithium vapor toward the graphite tiles of the lower center stack and divertor. Lithium oven depositions from a few milligrams to 1g have been applied between discharges. Benefits from the lithium coatings were sometimes, but not always, seen. These benefits sometimes included decreases in plasma density, inductive flux consumption, and edge-localized mode occurrence, and increases in electron temperature, ion temperature, energy confinement, and periods of edge and magnetohydrodynamic quiescence. In addition, reductions in lower divertor D, C, and O luminosi...

The influence of displacement damage on deuterium retention in tungsten exposed to plasma

Trapping of tritium at lattice damage from fusion neutron irradiation is expected to increase the tritium inventory in tungsten components in ITER. The magnitude of this increase depends on the concentration of traps that are produced, and on the depth to which the increased tritium retention extends into the material. Experiments to address these issues are described, in which displacement damage by ion irradiation was used as a surrogate for neutron damage. Irradiated samples were exposed to high flux deuterium plasma to simulate divertor conditions. The resulting deuterium content was measured by nuclear reaction analysis. Measurements were done at various damage levels up to those expected from the end-of-life neutron fluence in ITER. These experiments determine the number of traps produced by displacement damage and the rate at which they are filled during exposure to plasma. The role of defect annealing was explored through plasma exposures at various temperatures. In addition to trapping at damage, near-surface retention from internal precipitation was observed at lower temperatures. Addition of 5% helium to the deuterium plasma greatly reduced D retention by precipitation by localizing it closer to the surface. Results from these experiments indicate that the contribution to tritium inventory in ITER from trapping at neutron damage should be small.

First tests of molybdenum mirrors for ITER diagnostics in DIII-D divertor

Metallic mirrors will be used in ITER for optical diagnostics working in different spectral ranges. Their optical properties will change with time due to erosion, deposition, and particle implantation. First tests of molybdenum mirrors were performed in the DIII-D divertor under deposition-dominated conditions. Two sets of mirrors recessed 2cm below the divertor floor in the private flux region were exposed to a series of identical, lower-single-null, ELMing (featuring edge localized modes) H-mode discharges with detached plasma conditions in both divertor legs. The first set of mirrors was exposed at ambient temperature, while the second set was preheated to temperatures between 140 and 80°C. During the exposures mirrors in both sets were additionally heated by radiation from the plasma. The nonheated mirrors exhibited net carbon deposition at a rate of up to 3.7nm∕s and suffered a significant drop in reflectivity. Net carbon deposition rate on the preheated mirrors was a factor of 30–100 lower and their...

Hydrogen isotope retention in beryllium for tokamak plasma-facing applications

Beryllium has been used as a plasma-facing material to effect substantial improvements in plasma performance in the Joint European Torus (JET), and it is planned as a plasma-facing material for the first wall (FW) and other components of the International Thermonuclear Experimental Reactor (ITER). The interaction of hydrogenic ions, and charge-exchange neutral atoms from plasmas, with beryllium has been studied in recent years with widely varying interpretations of results. In this paper we review experimental data regarding hydrogenic atom inventories in experiments pertinent to tokamak applications and show that with some very plausible assumptions, the experimental data appear to exhibit rather predictable trends. A phenomenon observed in high ion-flux experiments is the saturation of the beryllium surface such that inventories of implanted particles become insensitive to increased flux and to continued implantation fluence. Methods for modeling retention and release of implanted hydrogen in beryllium are reviewed and an adaptation is suggested for modeling the saturation effects. The TMAP4 code used with these modifications has succeeded in simulating experimental data taken under saturation conditions where codes without this feature have not. That implementation also works well under more routine conditions where the conventional recombination-limited release model is applicable. Calculations of tritium inventory and permeation in the ITER FW during the basic performance phase (BPP) using both the conventional recombination model and the saturation effects assumptions show a difference of several orders of magnitude in both inventory and permeation rate to the coolant.

Effect of He on D retention in W exposed to low-energy, high-fluence (D, He, Ar) mixture plasmas

W targets are exposed at fixed temperature in the range ~420–1100 K, to either pure D2, D2–δHe (0.1 < δ < 0.25), or D2–δHe–γAr (γ = 0.03) mixture plasma, or He pretreatment plasma followed by exposure to D2 plasma. A strong reduction in D retention is found for exposure temperature above 450 K and incident He-ion fluence exceeding ~1024 m−2. Reduced D retention values lie well below that measured on D2 plasma-exposed reference targets, and the scatter in retention values reported in the literature. A small level of Ar admixture to D2–0.1He plasma, leading to an Ar ion density fraction of ~3%, is found to have minimal effect on the D inventory reduction caused by He. In targets with reduced inventory, nuclear-reaction analysis reveals shallow D trapping (<50 nm), in the same locale as nanometre-sized bubbles observed using transmission electron microscopy. It is suggested that near-surface bubbles grow and interconnect, forming pathways leading back to the plasma–material interaction surface, thereby interrupting transport to the bulk and reducing D retention.

Comparison of the thermal stability of the codeposited carbon/hydrogen layer to that of the saturated implant layer

Abstract This paper presents the results of an experimental study of the thermal stability in air and vacuum of the codeposited carbon/hydrogen layer formed in a carbon-lined fusion reactor. Results are also presented for the stability of the saturated layer formed by the implantation of energetic hydrogen ions into a graphite surface. For both films, the hydrogen isotope release occurs at a much lower temperature in air than it does in vacuum. At temperatures above 600 K, the hydrogen isotope release in air is very rapid and is emitted in a condensible form. It is speculated that isotopic exchange with water present in the air is responsible for this release. In vacuum, temperatures in excess of 1000 K are required to produce a rapid release from either film. The implications of these results to the safety of tritium in carbon-lined fusion reactors are discussed.

Trapping of deuterium at damage in graphite

Enhanced retention of deuterium (D) and tritium (T) in graphite due to neutron damage could result in large T inventories in future DT fueled fusion devices such as CIT and ITER. This paper describes experiments done to characterize the effect of lattice damage on retention of D in graphite. Lattice damage was produced in several types of graphite by irradiation with 6 MeV C+ ions or with neutrons. The damaged graphite was then heated to 1200{degree}C and exposed to D{sub 2} gas. The concentration of D retained in the graphite was measured by nuclear reaction analysis and compared to the level of damage. In POCO, N3M and H451 graphites, D retention increased with damage at low damage levels but saturated above 0.04 dpa at a D concentration of about 650 atomic ppM. D retention in damaged highly oriented pyrolytic graphite was an order of magnitude smaller than in the other graphites indicating that crystalline microstructure is an important factor. These results indicate that neutron damage in CIT and ITER may cause retention of large inventories of tritium in graphite components. 5 refs., 4 figs., 1 tab.

Hydrogen adsorption on and solubility in graphites

The experimental data on sorption and solubility of hydrogen isotopes in graphite in a wide ranges of temperature and pressure are reviewed. The Langmuir type adsorption is proposed for the hydrogen -- graphites interaction with taking into account dangling sp{sup 2}{minus}bonds relaxation. Three kinds of traps are proposed: Carbon interstitial loops with the adsorption enthalpy of {minus}4.4 eV/H{sub 2} (Traps l); carbon network edge atoms with the adsorption enthalpy of {minus}2.3 eV/H{sub 2} (Traps 2): Basal planes adsorption sites with enthalpy of +2.43 eV/H{sub 2} (Traps 3). The sorption capacity of every kind of graphite could be described with its own unique set of traps. The number of potential sites for the ``true solubility`` (Traps 3) we assume as 1E+6 appm, or HC=l, but endothermic character of this solubility leads to negligible amount of inventory in comparison with Traps 1 and Traps 2. The irradiation with neutrons or carbon atoms increases the number of Traps 1 and Traps 2. At damage level of {approximately}1 dpa under room temperature irradiation the number of these traps was increased up to 1500 and 5000 appm respectively. Traps 1 and Traps 2 are stable under high temperature annealing.