• Energy Research

Collector probes have long been used to measure impurity fluxes in the scrape off layer (SOL) of tokamaks. In this study, collector probes were inserted in the main SOL of DIII-D during the tungsten Metal Rings Campaign, and the W deposits on the probes were analyzed ex situ using Rutherford backscattering spectrometry analysis to obtain radial profiles of W deposition. A simple picture is hypothesized for how the W transports through the SOL from the target to the two sides of the probe, based on a long-theorized impurity accumulation at the crown (i.e. the top) of the plasma. The patterns observed in the deposition profiles along the probes support the hypothesized picture; however, 3D modeling and further experimental studies are needed to support more definitive conclusions.

The plasma surface interactions within a fusion device are one of the limiting factors to long pulse, power generating operation. A plasma facing component material will require effective heat tolerance, minimal erosion yield, and minimal fuel retention properties. Tungsten (W) has been selected as the divertor material for the International Thermonuclear Experimental Reactor (ITER) due to its high thermal conductivity and high sputter threshold. However, when W is exposed to high particle flux (>1022 ions/m2s) at high surface temperatures (>600 °C), the surface will develop defects such as pits, blisters, and nano-structured tendrils, reducing the beneficial properties of W. To overcome this limitation, a more radiation tolerant thin film material could be used, such as lithium (Li). In addition, the lithium film can protect the plasma from high-Z W sputtered atoms. In multiple tokamak devices, a Li wall coating, has improved the plasma performance by reducing fuel recycling from the walls, stabilizing the edge plasma and decreasing the number of edge localized modes (ELMs). Since ELMs help eject impurities from the core plasma, the complete suppression of ELMs is detrimental. Methods to regulate the frequency of ELMs have been investigated using gas puffs. In this work we report a new method to control the ELM frequency by tuning fuel recycling via the intrinsic helium (He) ions produced as ash from the deuterium (D) – tritium (T) fusion reaction. In Li films, one mechanism to retain D is via a chemical interaction between Li, O (oxygen), and D. Previous work has shown that when He ions are introduced with D ions, in a dual beam irradiation of Li films on W, a reduction in the dynamic surface D retention is observed. To further investigate this phenomenon, 1–2 um films of Li on W were exposed to sequential irradiations of D and He. The He fluence was ≈5% of the D (3.3 × 1020 ions/m2). The energies for the He and D ions were 1000 eV and 250 eV/amu, respectively and samples were exposed at room temperature. The surface chemistry was characterized with x-ray photoelectron spectroscopy (XPS) to determine changes in retention. The XPS scans were conducted in-situ and in-operando for the irradiations. Our results showed a decrease in the surface retention when He follows D ions and little change in the retention when D follows He. This indicates that He breaks the D retention mechanism in Li. Keywords: Lithium, Tungsten, Deuterium, In-situ XPS, Plasma facing components, Hydrogen retention

This work examines the effect of boron and lithium conditioned ATJ graphite surface bombarded by low-energy deuterium atoms on the deuterium retention and chemical sputtering. We use atomistic simulations and compare them with experimental in situ studies with x-ray photoelectron spectroscopy (XPS), to understand the effects of deuterium irradiation on the chemistry in lithiated, boronized and oxidized amorphous carbon surfaces. Our results are validated qualitatively by comparison with experiments and with quantum classical molecular dynamic simulations. We explain the important role of oxygen in D retention for lithiated surfaces and the suppression of oxygen role by boron in boronized surfaces. The calculated increase of the oxygen role in deuterium uptake after D accumulation in BCO surface configuration is discussed. The sputtering yield per low energy D impact is significantly smaller in boronized than in lithiated surfaces. Keywords: Plasma-surface interactions, Deuterium retention, NSTX, Boronization, Lithiumization, Carbon