R.G. Macaulay-Newcombe

Deuterium retention in single crystal tungsten

Abstract The retention of deuterium in single crystal tungsten (SCW) has been measured at 300 and 500 K, as a function of incident ion fluence over the range 10 21 –10 24 D + / m 2 . Irradiation of SCW with 1.5 keV D 3 + ions at 300 K leads to saturation at a much lower incident fluence than seen in polycrystalline tungsten (PCW), but with the same levels of D retention at saturation, ≈5×10 20 D / m 2 . Implantations at 500 K reached saturation at a very low incident fluence, below 10 21 D + / m 2 , with the amount of D retained at saturation ≈1.5×10 20 D / m 2 . This level is 3–4 times lower than the saturation value for 300 K implantation of the same single crystal of tungsten. Deuterium depth profile analysis by secondary ion mass spectrometry (SIMS) shows D trapping primarily within the 500 eV D + ion implantation range for both 300 and 500 K profiles. SIMS also revealed that the depth profiles for oxygen and deuterium were similar. When the tungsten was annealed at 500 K for 1 h after implantation at 500 K, SIMS indicated that the deuterium retention decreased by an order of magnitude.

Chemical erosion of DIII-D divertor tile specimens

Abstract Chemical erosion measurements have been performed on graphite tile specimens taken from the upper and lower divertors of the DIII-D tokamak. Measurements for 50 eV/D+ incident on the lower divertor resulted in erosion yields which are the same as those for graphite. This implies that any deposited film on the tile surface was completely removed after very short beam exposures (∼1021 D+/m2). For the upper divertor tiles, the measured erosion yields for 50 and 200 eV/D+ were somewhat larger than those for graphite at 300 and 500 K; however, at 700 K the yields were lower than graphite yields and were more comparable to boron-doped graphite. Thermo-oxidation of both upper and lower divertor specimens also indicated the rapid removal of surface films, followed by the slower removal of implanted deuterium from the graphite substrate.

Chemical erosion of boronized films from DIII-D tiles

Abstract Tile specimens from the DIII-D tokamak have been studied to determine their erosion characteristics when exposed to D+ ions and O2 gas. Here, we report results for tile surfaces from the outer midplane. Surface analyses (EDX, XPS, SIMS) indicate that the surface layer is composed primarily of boron, with an overlayer of a B/C mixture. Total hydrocarbon (ΣCiDj) erosion yields were initially found to be ∼0.01–0.02 C/D+, with limited variations due to D+ energy (50 or 200 eV) or specimen temperature (300–700 K). Erosion yields were seen to decrease with fluence by a factor of 1.5–2 over the range ∼4×1021–3×1022 D+/m2. Above ∼ 3×10 22 D + / m 2 the yields level off. The initial erosion yields are found to be consistent with those for boron-doped graphite. O2 gas exposure at 523 or 623 K initially removed ∼25% of the trapped D; however, the remaining D could not be removed by baking in O2 at temperatures up to 623 K.

Thermo-oxidation of hard carbon films with tungsten surface impurity

Abstract The removal of codeposited tritiated carbon films from the next generation of fusion reactors may involve baking in an O 2 environment. Experimental results have indicated that thermo-oxidation can be effective in the removal of such films, however, wide variations have been observed in the oxidation rates of various types of carbon films. In the current experiments, we have investigated the role of metallic impurities by sputter-depositing tungsten onto hard a-C:D films, and exposing them to O 2 gas at 623 K. It was found that rather than catalysing the oxidation of the hydrogenated carbon film, the W deposit tended to inhibit the film removal at this temperature. This suggests that film structure is the predominant factor determining the oxidation rate of tokamak codeposits.

An Interpretation Of The Retention Of Low Energy Deuterium Ions In Tungsten

This report is a summary of our investigations of D retention in various forms of tungsten: high purity polycrystalline W (PCW), polycrystalline W containing 1% La2O3, and two grades of single crystal W (SCW). The experiments have been primarily implantations followed by thermal desorption measurements. Nuclear reaction analysis and SIMS were used to measure the depth distributions near the surface. By using low energy ions (500 eV/D+), it has been possible to eliminate most if not all elastic collision defect creation, yet this has not reduced the trapping significantly. Comparing the results for the various forms of W has enabled us to assess the effects of grain boundaries, dislocations and impurities. Recent work has included a study of the effects of ion flux on deuterium retention in SCW. The conclusion drawn is that D is trapped in clusters or nano-bubbles, and that these traps grow with increasing fluence, particularly at higher temperatures. The large variations in the retention of D in W show that the processes controlling D trapping are many and complex. In order to form a more comprehensive picture, and eventually attempt to predict retention under fusion relevant conditions, modeling has been performed using TMAP4[1].