Fundamental processes of plasma and reactive gas surface treatment for the recovery of hydrogen isotopes from carbon co-deposits in fusion devices
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Beschreibung
The use of carbon-based plasma-facing wall components offers many advantages for plasma operation
in magnetic confinement nuclear fusion devices. However, through reactions with the hydrogen based
fusion 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. CO2 and H2O. These gases are pumped from the vacuum vessel and recycled. Applying the
model to literature observations enables to connect data on exposure temperature, pressure, time and
co-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 plasma
sheath 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 O2 and NO2, at temperatures of 470 to 630K and pressures of
2 and 20kPa 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 Rutherfordbackscattering-
spectrometry (NRA/RBS). Layer thicknesses are measured with ellipsometry. The
experiments are conducted using well defined, high purity a-C:D layers deposited by plasmachemical-
vapour-deposition from CD4 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 NO2–TCR at 630K is possible
within 3min. 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 D2 with a removal rate of 5.7±0.9*1015 D/(cm²s) as the fastest base gas in tokamaks.
Comparisons with O2 show that the higher sputtering yield of O is counteracted by an 11-fold lower
ion surface flux density, introduced by fundamental properties of O2. Pumping speed and partial
exhaust gas pressures are identified as limiting factors for the removal rate, explaining differences to
non-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 D2-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 a
tritium inventory of 700g to 10.7h in an application scenario
Autor*in
Sören Möller