Surface Modification of ITER-like Mirrors after One Hundred Cleaning Cycles Using Radio-Frequency Plasma

In ITER, the metallic first mirrors (FMs) will undergo erosion due to their proximity to the fusion plasma and deposition of materials originated from the first walls (mainly beryllium). In-situ plasma cleaning is a promising technique to conserve the FMs optical properties by means of ion sputtering. In this work, the evolution of the optical properties of single-crystal (Sc) and nanocrystalline (Nc) molybdenum (Mo) and rhodium (Rh) mirrors were investigated up to 100 cycles of consecutive contamination and cleaning. Aluminum oxide (AlO) was used as contaminant to replace the toxic beryllium. The plasma cleaning was carried out using a capacitively coupled argon (Ar) plasma excited by a 60 MHz radio-frequency generator resulting in the formation of a self-bias applied on the mirrors of -280 V. The plasma potential being around 30 V, the Ar ion energy was about 310 eV. The optical properties of the mirrors were assessed using ex-situ reflectivity measurements. Moreover, the surface topography was characterized by means of scanning electron microscopy (SEM), focused ion beam (FIB) and roughness measurements using atomic force microscopy (AFM). ScMo and ScRh mirrors formerly exposed to 80 successful cleaning cycles using aluminum/tungsten (Al/W) deposits and air storage exhibit drastic changes in their optical properties after being subject to cleaning cycles using AlO as contaminant. Additionally, freshly polished ScRh were exposed to identical cleaning cycles. All Sc mirrors exhibited pits induced by the polishing procedure using diamond paste in addition of mounds/wavy patterns. The carbon incorporated during the polishing process was demonstrated to be responsible for the pitting of the surface. The Nc mirrors preserved their initial reflectivities after up to 100 cycles. The surface topography was systematically characterized and an average erosion rate for NcRh mirrors of about 59 nm per cycle has been estimated from FIB cross-sections. The optical properties of the Nc mirrors showed a superiority in the present study in comparison to the Sc materials due to the influence of their polishin

A randomised phase 2 study comparing different dose approaches of induction treatment of regorafenib in previously treated metastatic colorectal cancer patients (REARRANGE trial)

Altres ajuts: Bayer HealthCare Pharmaceuticals Inc.Purpose: The purpose of this article is to evaluate the safety of two regorafenib dose-escalation approaches in refractory metastatic colorectal cancer (mCRC) patients. Patients and methods: Patients with mCRC and progression during or within 3 months following their last standard chemotherapy regimen were randomised to receive the approved dose of regorafenib of 160 mg QD (arm A) or 120 mg QD (arm B) administered as 3 weeks of treatment followed by 1 week off, or 160 mg QD 1 week on/1 week off (arm C). The primary end-point was the percentage of patients with G3/G4 treatment-related adverse events (AEs) in each arm. Results: There were 299 patients randomly assigned to arm A (n = 101), arm B (n = 99), or arm C (n = 99); 297 initiated treatments (arm A n = 100, arm B n = 98, arm C n = 99: population for safety analyses). G3/4 treatment-related AEs occurred in 60%, 55%, and 54% of patients in arms A, B, and C, respectively. The most common G3/4 AEs were hypertension (19, 12, and 20 patients), fatigue (20, 14, and 15 patients), hypokalemia (11, 7, and 10 patients), and hand-foot skin reaction (8, 7, and 3 patients). Median overall survival was 7.4 (IQR 4.0-13.7) months in arm A, 8.6 (IQR 3.8-13.4) in arm B, and 7.1 (IQR 4.4-12.4) in arm C. Conclusions: The alternative regorafenib dosing schedules were feasible and safe in patients with mCRC who had been previously treated with standard therapy. There was a higher numerical improvement on the most clinically relevant AEs in the intermittent dosing arm, particularly during the relevant first two cycles. Clinicaltrials.gov identifier: NCT02835924

A high power laser facility to conduct annealing tests at high temperature

International audienceThe knowledge of materials properties and their behavior at high temperatures is of crucial importance in many fields. For instance annealing phenomena occuring during the thermomechanical processing of materials, such as recrystallization, have long been recognized as being both of scientific interest and technological importance. Different methods are currently used to study annealing phenomena and submit metals to heat loads. In this work, we present the design and the development of a laser-based facility for annealing tests. This experimental setup enables studies at the laboratory scale with great flexibility to submit samples to various spatial and temporal heating profiles. Due to the possibility to have optical access to the sample, laser heating can be combined to several non-contact diagnostics such as infrared imaging to control and analyse the temperature gradients. As case study, we present a set of experiments performed to study the recrystallization kinetics of tungsten. We demonstrate that samples can be heated linearly with heating rate up to ∼2000 K/s, at temperatures above 2000K, for seconds or hours, with typical errors in the temperature measurement of around 1% that depend mainly on the determination of sample emissivity. Such studies are of crucial interest in the framework of nuclear fusion, since the ITER nuclear reactor will operate with a full-W divertor

Material erosion, deposition and material transport in the stellarator W7-X

Material erosion, deposition and material transport in the stellaratorW7-XM. Mayera,*, M. Baldena, S. Brezinsekb, V.V. Burwitza,c, C.P. Dhardd, R. Guimarãese, M. Guitart Corominasa, P. Hireta, D. Naujoksd, R. Neua,c, J.H. Schmidt-Denckera, T.S. Silvae, and W7-X TeamaMax-Planck-Institut für Plasmaphysik, Garching, GermanybForschungszentrum Jülich GmbH, Jülich, GermanycTechnische Universität München, München,GermanydMax-Planck-Institut für Plasmaphysik, Greifswald, GermanyeUniversity of São Paulo, São Paulo, BrazilNet erosion, deposition and material transport in the stellarator W7-X were investigated on the Test Divertor Unit (TDU) using special carbon marker coatings and on the vessel walls by analysis of W-coated and regular wall elements duringthe operational phases OP1.2a in the year 2017 and OP1.2b in 2018. While OP1.2a was characterised by high concentrations of oxygen and carbon impuritiesin the plasmas, OP1.2b had much smaller impurity concentrations due to regular boronizations and showed considerably better plasma performances with higher plasma densities. First exposures of tungsten marker coatings in the TDU and at the inner heat shield were performed in OP1.2b. The samples were analysed by quantitative ion beam analysis methods, scanning electron microscopy, and laser-induced breakdown spectroscopy (LIBS).Very high net erosion of carbon was observed at the strike line of all 10 TDUs in both campaignsand is attributed to sputtering and chemical erosion by C and O impurities in OP1.2a[1], while the erosion was probably dominated by impact of H ions in OP1.2b. Re-deposition of carbon in remote areas of the TDU was small.Thicker re-deposited carbon layers with thicknesses of a few microns were observed at the divertor baffles,anoticeable deposition of boron on the TDU was observed after OP1.2b. The inner wall showed no erosion but net deposition of about 100 nm thick boron/carbon layersdue to the boronizations, the outer vessel wall showed a complicated pattern of small erosion areas and thin deposits. Stripes of boron/carbon layers were observed behind the gaps of inner wall protection tiles.The global carbon balance is currently unclear. However, carbon eroded from the TDU was not redeposited in the divertor region but was transported out of the divertor area. Some carbon was redeposited at the divertor baffles but was also pumped out as CO, CO2, or CH4. This is a profound difference to divertors in tokamaks, where eroded material is typically redeposited in remote divertor areas or in the inner divertor. A first assessment of tungsten as plasma-facing material in W7-X is provided.[1] M. Mayer et al., PFMC 2019, Phys. Scr., in press*Corresponding author:tel.: +49 89 3299 1639, e-mail: [email protected]