Fundamental processes of plasma and reactivegas surface treatment for the recovery ofhydrogen isotopes from carbon co-deposits infusion devices

The use of carbon-based plasma-facing wall components offers many advantages for plasma operationin magnetic confinement nuclear fusion devices. However, through reactions with the hydrogen basedfusion plasma, carbon forms amorphous hydrogenated carbon co-deposits (a-C:H) in the vacuum vessels. If tritium is used to fuel the reactor, this co-deposition can quickly lead to an inacceptable high tritium inventory. Through co-deposition with carbon about 10% of the tritium injected into the reactor can be trapped. Even with other wall materials co-deposition can be significant. A method to recover the hydrogen isotopes from the co-deposits is necessary. The method has to be compatible with the requirements of the devices and nuclear fusion plasma operation. In this work thermo-chemical removal by neutral gases (TCR) and removal by plasmas is investigated. Models are developed to describe the involved processes of both removal methods. TCR is described using a reaction-diffusion model. Within this model the reactive gas diffuses into the co-deposits and subsequently reacts in a thermally activated process. The co-deposits are pyrolysed, forming volatile gases, e.g. CO$_{2}$ and H$_{2}$O. These gases are pumped from the vacuum vessel and recycled. Applying themodel to literature observations enables to connect data on exposure temperature, pressure, time andco-deposit properties. Two limits of TCR (reaction- or diffusion-limited) are identified. Plasma removal sputters co-deposits by their chemical and physical interaction with the impinging ions. The description uses a 0D plasma model from the literature which derives plasma parameters from the balance of input power to plasma power losses. The model is extended with descriptions of the plasmas heath and ion-surface interactions to derive the co-deposit removal rates. Plasma removal can be limited by this ion induced surface release rate or the rate of pumping of the released species. To test the models dedicated experiments are conducted. Sets of a-C:D layers with different thickness and structure are exposed to TCR, using O$_{2}$ and NO$_{2}$, at temperatures of 470 to 630K and pressures of 2 and 20 kPa to investigate the strong impact of exposure and layer properties, as predicted by the model. Plasmas produced by electron (ECR) and ion cyclotron frequencies (ICWC) are investigated with several base gases in a compact toroidal plasma device and the tokamak TEXTOR. The ion fluxes of these plasmas are investigated with Langmuir probes to allow the model comparison. Pre/Post determination of the layers allows quantifying the removal rates of the tested methods. The areal density of deuterium and carbon is determined by nuclear reaction analysis and Rutherford backscattering-spectrometry (NRA/RBS). Layer thicknesses are measured with ellipsometry. The experiments are conducted using well defined, high purity a-C:D layers deposited by plasma chemical-vapour-deposition from CD$_{4}$ in a specifically adapted vacuum device to be able to separate the effects of layers properties and exposure parameters. The experiments demonstrate that a 95% removal of a-C:D layers with NO$_{2}$–TCR at 630K is possible within 3 min. The model´s prediction of a linear relation between the TCR rate and the co-deposits inventory is experimentally approved, validating its volume effect. The experiments with plasma removal reveal D$_{2}$ with a removal rate of 5.7 ± 0.9 $^{*}$ 10$^{15}$ D/(cm$^{2}$s) as the fastest base gas in tokamaks. Comparisons with O$_{2}$ show that the higher sputtering yield of O is counteracted by an 11-fold lowerion surface flux density, introduced by fundamental properties of O$_{2}$. Pumping speed and partial exhaust gas pressures are identified as limiting factors for the removal rate, explaining differences tonon-local observations from the literature. Furthermore, it is possible to remove O stored in surfaces in TEXTOR in, for fusion plasma operation, detrimental amounts by D$_{2}$-ICWC. The models are in agreement with literature and new experimental data obtained in this work. Using the new knowledge, the methods can be adapted to future devices, e.g. ITER. TCR offers a fast removal with only logarithmic scaling with co-deposit inventory, while plasma removal results in good wall conditions for fusion operation. The proposed integral scenario combines both specific advantages to a fusion plasma compatible removal scenario. The determined removal rates and the technical specifications of ITER are used to calculate the removal time at 470K wall temperature for atritium inventory of 700g to 10.7h in an application scenario