Search

If both carbon and tungsten were to be part of the plasma-facing armor in a future fusion reactor, it is inevitable that carbon co-deposits containing tungsten impurities will form. This work examines the effectiveness of thermo-oxidation in removing hydrogen from W-containing carbon co-deposits. Amorphous deuterated hydrocarbon (a-C:D) films were created with a CD4/Ar direct-current glow discharge and doped with W sputtered from a W mesh in front of the specimen. The W concentration in the specimens ranged from 0 to 35 at. % W/(W + C). The films were oxidized at 350°C, in 2 Torr pure O2 for time increments totaling 8 h. The D content of the films was measured before and at various stages of the oxidation exposure using laser thermal desorption spectroscopy. Essentially all deuterium was removed from films containing very little or no W doping [

Thermo-oxidation is a technique where O2 gas is used to remove carbon co-deposits at elevated temperatures from plasma-facing materials of fusion devices. Following thermo-oxidation, the reaction products produced will need to be managed by a reactor's gas-handling system, including the tritium plant. Thus, it is important to know what molecular species are produced during oxidation. The reaction products of thermo-oxidation have previously been studied for undoped films indicating CO, CO2 and D2O as the reaction products. The present study directly compares the reaction product evolution of undoped films to that of films doped with tungsten.A glow discharge was used to produce sets of undoped and W-doped films on stainless steel (SS) foil substrates. X-ray photoelectron spectroscopy (XPS) found the W-doped films to contain an average 0.1 at.% W/(W+C). Laser thermal desorption spectroscopy (LTDS) was used to measure the areal D-concentration. Reaction products during oxidation were measured using a quadrupole mass spectrometer (QMS). Both the undoped and W-doped specimens were oxidized in 2Torr O2, at 350 °C for 4h, total. LTDS was then used to determine how much D had been removed during oxidation (93% and 32% of the D was removed after oxidation for the undoped and W-doped cases respectively). The reduced D loss for the W-doped specimens is attributed to a change in film structure due to the incorporation of W in the film. Oxidation reaction products for both undoped and W-doped films include CO, CO2, and D2O, with no finding of D2 or CD4 within experimental uncertainty. Particle accounting was performed for oxygen, deuterium and carbon atoms. Keywords: Plasma-surface interactions, Tritium, Carbon, Carbon co-deposition, Oxidation, Reaction products

As a candidate accident tolerant fuel cladding material, the high temperature oxidation behaviour of an Oxygen Dispersion Strengthened (ODS) Fe–12Cr–6Al alloy has been examined in isothermal air and steam environments. While the oxidation behaviour of FeCrAl has been extensively studied, very little has been reported in the literature on FeCrAl-ODS, which is the focus of this manuscript. By conducting Thermogravimetric Analysis (TGA), oxidation experiments were performed on specimens of Fe–12Cr–6Al-ODS at temperatures up to 1400 ∘C in air and up to 1215 ∘C in high purity steam. Oxidation kinetics were evaluated by calculating reaction rate constants and activation energies. The formation of an alumina scale was correlated to kinetics results from TGA by coupling Scanning Electron Microscopy (SEM) with Energy Dispersive X-ray Spectroscopy (EDS) line scans to investigate the chemical composition of the primary oxide layer, local oxidation effects, as well as the bulk alloy material. Electron Probe Micro-Analysis (EPMA) point scans were applied to the heavily oxidized specimen to achieve an accurate composition profiling of areas of interest. Finally, thermodynamic calculations were performed to interpret the formation of the complex oxide layer. These experiments support research efforts for the development of accident tolerant fuel technologies at Nippon Nuclear Fuel Development Co., Ltd.

Thermo-oxidation is controlled exposure in anoxygen-containing atmosphere at elevated temperatureand is being considered as a technique for the de-tritiation of carbon-based codeposits in ITER. In addi-tion, unplanned oxidation may also occur duringaccidental air ingress. The impact of thermo-oxidationon ITER diagnostic mirrors causes concerns. A dedi-cated study was performed in DIII-D, where molybde-num and copper mirrors were installed in the mainchamber, in the divertor, and at a location remote fromthe plasma and exposed for

This work extends previous thermo-oxidation studies [R. Ochoukov, A.A. Haasz, J.W. Davis, Phys. Scr. T124 (2006) 27; A.A. Haasz, C.K. Tsui, J.W. Davis, R. Ochoukov, Phys. Scr. T128 (2007) 55] of DIII-D codeposits to provide reference data prior to performing in situ thermo-oxidation in DIII-D. The present results confirm that codeposit removal by oxidation in dry air or a mixture of O2–He gas is equally effective as in pure O2, given equivalent O2 partial pressures. We have also demonstrated that increasing the oxidation time to 48 h (compared with 2–8 h [R. Ochoukov, A.A. Haasz, J.W. Davis, Phys. Scr. T124 (2006) 27; A.A. Haasz, C.K. Tsui, J.W. Davis, R. Ochoukov, Phys. Scr. T128 (2007) 55]) does not compensate for ‘less severe’ oxidation conditions of (i) ‘low pressure’ (0.021 kPa O2 partial pressure) at ‘high temperature’ (623 K) and (ii) ‘low temperature’ (523 K) at ‘high pressure’ (17 kPa O2 partial pressure). The new results also show that thermo-oxidation is equally effective in removing codeposits from open surfaces and from tile gaps accessible by narrow (∼1 mm) passageways, and have extended the range of the linear relationship between D removal rates and inherent D content.

The oxidation of dust and flakes collected from the DIII-D tokamak, and various commercial dust specimens, has been measured at 350 °C and 2.0 kPa O 2 pressure. Following an initial small mass loss, most of the commercial dust specimens showed very little effect due to O 2 exposure. Similarly, dust collected from underneath DIII-D tiles, which is thought to comprise largely Grafoil™ particulates, also showed little susceptibility to oxidation at this temperature. However, oxidation of the dust collected from tile surfaces has led to ∼18% mass loss after 8 h; thereafter, little change in mass was observed. This suggests that the surface dust includes some components of different composition and/or structure – possibly fragments of codeposited layers. The oxidation of codeposit flakes scraped from DIII-D upper divertor tiles showed an initial 25% loss in mass due to heating in vacuum, and the gradual loss of 30–38% mass during the subsequent 24 h exposure to O 2 . The oxidation of these flakes was much slower than that observed for the oxidation of thinner DIII-D codeposit specimens which were still adhered to tile surfaces. This is thought to be related to structural differences.

In April 2010, two thermo-oxidation experiments (?O-bakes?) were performed in the DIII-D tokamak. Internal surfaces of the tokamak, as well as a number of specimens inserted into the torus, were exposed to a mixture of 20% O2/80% He at a nominal pressure of 9.5?Torr (1.27?kPa) at a temperature of 350?360??C for a duration of 2?h. Three primary conclusions have been drawn from these experiments: (1) laboratory measurements on the release of deuterium from tokamak codeposits by oxidation have been duplicated in a tokamak environment, (2) no internal tokamak components or systems were adversely affected by the oxidation and (3) the recovery of plasma performance following oxidation was similar to that following regular torus openings.

An experiment was conducted in DIII-D to examine carbon deposition when a secondary separatrix is near the wall. The magnetic configuration for this experiment was a biased double-null, similar to that foreseen for ITER. 13C methane was injected toroidally symmetrically near the secondary separatrix into ELMy H-mode deuterium plasmas. The resulting deposition of 13C was determined by nuclear reaction analysis. These results show that very little of the injected 13C was deposited at the primary separatrix, whereas a large fraction of injected 13C was deposited close to the point of injection near the secondary separatrix. Six of the tiles were put back into DIII-D, where they were baked at 350–360 °C for 2 h at ~1 kPa in a 20% O2/80% He gas mixture. Subsequent ion beam analysis of these tiles showed that about 21% of the 13C and 54% of the deuterium were removed by the bake.