Polymers for O-rings, valve seats, gaskets, and other sealing applications in the hydrogen infrastructure face extreme conditions of high-pressure H2 (0.1 to 100 MPa) during normal operation. To fill current knowledge gaps and to establish standard test methods for polymers in H2 environments, these materials can be tested in laboratoryscale H2 manifolds mimicking end use pressure and temperature conditions. Beyond the influence of high pressure H2, the selection of gases used for leak detection in the H2 test manifold, their pressures and times of exposure, gas types, relative diffusion and permeation rates are all important influences on the polymers being tested. These effects can be studied ex-situ with post-exposure characterization. In a previous study, four polymers (Viton A, Buna N, High Density Polyethylene (HDPE) and Polytetrafluoroethylene (PTFE)), commonly used in the H2 infrastructure, were exposed to high-pressure H2 (100 MPa). The observed effects of H2 were consistent with typical polymer property-structure relationships; in particular, H2 affected elastomers more than thermoplastics. However, since high pressure He was used for purging and leak detection prior to filling with H2, a study of the influence of the purge gas on these polymers was considered necessary to isolate the effects of H2 from those of the purge gas. Therefore, in this study, Viton A, Buna N, and PTFE were exposed to the He purge procedure without the subsequent H2 exposure. Additionally, six polymers, Viton A, Buna N, PTFE, Polyoxymethylene (POM), Polyamide 11 (Nylon), and Ethylenepropylenediene monomer rubber (EPDM), were subjected to high pressure Ar (100 MPa) followed by high pressure H2 (100 MPa) under the same static, isothermal conditions to identify the effect of a purge gas with a significantly larger molecular size than He. Viton A and Buna N elastomers are more prone to irreversible changes as a result of H2 exposure from both Ar and He leak tests as indicated by influences on storage modulus, extent of swelling, and increased compression set. EPDM, even though it is an elastomer, is not as prone to high-pressure gas influences. The thermoplastics are generally less influenced by high pressure regardless of the gas type. Conclusions from these experiments will provide insight into the influence of purging processes and purge gases on the subsequent testing in high pressure gaseous H2. Control for the influence of purging on testing results is essential for the development of robust test methods for evaluating the effects of H2 and other high-pressure gases on the properties of polymers.

In this work, we examine the response of an ultra-fine grained (UFG) tungsten material to high-flux deuterium plasma exposure. UFG tungsten has received considerable interest as a possible plasma-facing material in magnetic confinement fusion devices, in large part because of its improved resistance to neutron damage. However, optimization of the material in this manner may lead to trade-offs in other properties. We address two aspects of the problem in this work: (a) how high-flux plasmas modify the structure of the exposed surface, and (b) how hydrogen isotopes become trapped within the material. The specific UFG tungsten considered here contains 100 nm-width Ti dispersoids (1 wt%) that limit the growth of the W grains to a median size of 960 nm. Metal impurities (Fe, Cr) as well as O were identified within the dispersoids; these species were absent from the W matrix. To simulate relevant particle bombardment conditions, we exposed specimens of the W-Ti material to low energy (100 eV), high-flux (> 1022 m− 2 s− 1) deuterium plasmas in the PISCES-A facility at the University of California, San Diego. To explore different temperature-dependent trapping mechanisms, we considered a range of exposure temperatures between 200 °C and 500 °C. For comparison, we also exposed reference specimens of conventional powder metallurgy warm-rolled and ITER-grade tungsten at 300 °C. Post-mortem focused ion beam profiling and atomic force microscopy of the UFG tungsten revealed no evidence of near-surface bubbles containing high pressure D2 gas, a common surface degradation mechanism associated with plasma exposure. Thermal desorption spectrometry indicated moderately higher trapping of D in the material compared with the reference specimens, though still within the spread of values for different tungsten grades found in the literature database. For the criteria considered here, these results do not indicate any significant obstacles to the potential use of UFG tungsten as a plasma-facing material, although further experimental work is needed to assess material response to transient events and high plasma fluence.

The absorption of CO2 laser radiation in p-type GaAs is dominated by direct free-hole transitions between states in the heavy- and light-hole bands. For laser intensities on the order of 10 MW/cm2, we report that the absorption associated with these transitions in moderately Zn-doped GaAs (∼1017 cm−3) begins to saturate in a manner predicted by an inhomogeneously broadened two-level model. At higher laser intensities surface melting occurs initially at localized sites in moderately doped material and more uniformly in heavily Zn-doped samples (≳1018 cm−3). As the energy density of the CO2 laser radiation is progressively increased further, the surface topography of the samples shows signs of ripple patterns, high local stress, vaporization of material, and exfoliation of solid GaAs fragments. Electron-induced x-ray emission data taken on the laser-melted samples show that there is a loss of As, compared to Ga, from the surface during the high-temperature cycling. By irradiating the samples in air, argon, and vacuum, we find that the vaporization rates are directly influenced by the ambient environment, particularly by the interaction of oxygen with the molten GaAs. Secondary ion mass spectrometry measurements are used to study the diffusion of oxygen from the native oxide and the incorporation of oxygen in the near-surface region of the GaAs samples that have been melted by a CO2 laser pulse. We find that oxygen incorporation does occur, and that the amount and depth of the oxygen incorporation depends on the laser energy density, number of laser shots, and ambient environment. For samples that are irradiated in argon or vacuum, we find that removal of the native oxide can be accomplished with CO2 laser pulses. Similar measurements are performed on Si-implanted GaAs, and results are reported for the redistribution of the implanted silicon atoms, the deviations from stoichiometry, and the incorporation of oxygen in the resolidified layer.

The Sustained Spheromak Physics Experiment (SSPX) will examine the confinement properties of spheromak plasmas sustained by DC helicity injection. Understanding the plasma-surface interactions is an important component of the experimental program since the spheromak plasma is in close contact with a stabilizing wall (flux conserver) and is maintained by a high current discharge in the coaxial injector region. Peak electron temperatures in the range of 400 eV are expected, so the copper plasma facing surfaces in SSPX have been coated with tungsten to minimize sputtering and plasma contamination. Here, we report on the characterization and conditioning of these surfaces used for the initial studies of spheromak formation in SSPX. The high pressure plasma-sprayed tungsten facing the SSPX plasma was characterized in situ using β-backscattering and ex situ using laboratory measurements on similarly prepared samples. Measurements showed that water can be desorbed effectively through baking while the removal rates of volatile impurity gases during glow discharge and shot conditioning indicated a large source of carbon and oxygen in the porous coating.

Thermal response and thermal fatigue tests of four 5-mm-thick beryllium tiles on a Russian Federation International Thermonuclear Experimental Reactor (ITER)-relevant divertor mock-up were completed on the electron beam test system at Sandia National Laboratories. The beryllium tiles were diffusion bonded onto an oxygen-free high-conductivity copper saddle-block and a dispersion-strengthened copper alloy tube containing a copper porous coating. Thermal response rests were performed on the files to an absorbed heat flux of 5 MW/m(2) and surface temperatures near 300 degrees C using 1.4 MPa wafer at 5 m/s flow velocity and an inlet temperature of 8 to 15 degrees C. One tile was exposed to incrementally increasing heat fluxes up to 9.5 MW/m(2) and surface temperatures up to 690 degrees C before debonding at 10 MW/m(2). A second tile debonded in 25 to 30 cycles at <0.5 MW/m(2). However, a third file debonded after 9200 thermal fatigue cycles at 5 MW/m(2), while another debonded after 6800 cycles. Posttest surface analysis indicated that fatigue failure occurred in the intermetallic layers between the beryllium and copper. No fatigue cracking of the bulk beryllium was observed. It appears that microcracks growing at the diffusion bond produced the observed gradual temperature increases during thermal cycling. These experiments indicate that diffusion-bonded beryllium files can survive several thousand thermal cycles under ITER-relevant conditions. However, the reliability of the diffusion-bonded joint remains a serious issue.

The chemical compatibility of various metallic and organic containment materials with a constituent of a paste extrudable explosive (PEX) has been examined through a series of long-term exposures. Corrosion coupons and mechanical test specimens (polymers only) were exposed to FM-1, a principal liquid component of PEX, at 74{degree}C. RX-08-FK is the LLNL designator for this formulation. Compatibility was determined by measuring changes in weight, physical dimensions, and mechanical properties, by examining the coupons for discoloration, surface attack, and corrosion products, and by analyzing for dissolved metals in the FM-1. Of the metals and alloys examined, none of the 300 series stainless steels exhibited adequate corrosion resistance after 74 days of exposure. Copper showed evidence of severe uniform surface attack. Monel 400 also exhibited signs of chemical attack. Nickel and tantalum showed less evidence of attack, although neither, was immune to the liquid. Gold coupons developed a ``tarnish`` film. The gold along with an aluminum alloy, 6061 (in the T6 condition) performed the most satisfactorily. A wide range of polymers were tested for 61 days at 74{degree}C. The materials that exhibited the most favorable response in terms of weight change, dimensional stability, and mechanical properties were Kalrez, PTFE Teflon, and polyethylene.

The plasma–material interaction and high heat flux properties of beryllium are reviewed to determine its suitability as a plasma‐facing component in magnetic fusion energy reactors. Consideration is given to beryllium’s outgassing, erosion, and hydrogen retention characteristics. Its responses to normal and off‐normal high heat fluxes are compared to graphite in both the as‐received and the neutron‐irradiated states. Beryllium’s performance in present‐day devices is assessed, and its expected behavior in future reactors is summarized. It is concluded that beryllium is potentially a better plasma‐facing material than graphite and that more development and testing is warranted.

Since the introduction of new graphite tiles on the divertor regions and the center post of DIII-D, beta backscatter measurements have been made of the metals codeposited on these surfaces at three different times. At the first vent (June 1988), the upper and lower portions of the machine appeared very similar, with the average amount of metal on a surface increasing with increasing radius and indications that local tile alignment had an effect on the amount of metal deposited. Measurements made during subsequent vents show that there has been much more metal deposited on the lower divertor than on the upper. Metal seems to be concentrated at radii both less than and greater than that expected for the strike zones. From measurements made during the latest vent (December 1989) there is clear indication that material has been eroded from a position of the outer strike zone, probably to be redeposited at a larger radius.

During 1989, following a brief all carbon phase, JET was operated with an evaporated beryllium film on all the plasma facing surfaces and then, in a third phase, with bulk beryllium used for the belt limiters and one set of antenna protection tiles. Subsequent analysis of wall components and long term samples (LTS) using Beta Backscatter and Ion Beam Analysis has been conducted to determine the condition of the First Wall and its deuterium inventory after beryllium operation.Ex-situ analysis of components where deposition occurs during operation shows mixed carbon and beryllium layers with an approximate ratio 1:1 and some localised higher Z components. The deuterium inventory situation is littled changed by the introduction of beryllium. Similar surface levels of deuterium, approximately 1 x 10(22) atoms m-2, have been seen for both carbon and beryllium belt limiter tiles.