R. Mateus

Hydrogenic retention of high-Z refractory metals exposed to ITER divertor-relevant plasma conditions

Tungsten (W) and molybdenum (Mo) targets are exposed to the plasma conditions expected at the strike point of a detached ITER divertor (ne ~ 1020 m−3, Te ~ 2 eV) in the linear plasma device Pilot-PSI. The peak surface temperatures of the targets are ~1600 K for W and ~1100 K for Mo. The surface temperatures and plasma flux densities decrease radially towards the edges of the target due to the Gaussian distribution of electron density (ne) and temperature (Te) in the plasma column. A 2D spatial scan of the W and Mo targets using nuclear reaction analysis (NRA) shows D retention is strongly influenced by surface temperature in the range 800–1600 K and this dependence dominates over any plasma flux dependence. NRA and thermal desorption spectroscopy (TDS) show no clear dependence of retention on incident plasma fluence for the W targets with retained fractions ranging from 10−8–10−5 Dretained/Dincident. NRA and TDS for the Mo targets show retention rates a factor of 4–5 higher than the W targets and this is likely due to the lower surface temperatures for the Mo plasma exposures. NRA also reveals a thin boron layer on the Mo targets but the presence of boron does not correspond to a significant increase in D retention. Overall hydrogenic retention in W and Mo is shown to be low (Dretained = 1019–1020 D m−2) despite exposure to high plasma flux densities (~1024 D m−2 s−1). This is likely due to the elevated surface temperature due to plasma thermal loading during exposure.

Consolidation of Cu-nDiamond Nanocomposites: Hot Extrusion vs Spark Plasma Sintering

Due to their interesting properties copper-based materials have been considered appropriate heat-sinks for first wall panels in nuclear fusion devices. The concept of property tailoring involved in the design of metal matrix composites has led to several attempts to use nanodiamond (nDiamond) as reinforcement. In particular, nDiamond produced by detonation has been used to reinforce copper. In the present study, powder mixtures of copper and nDiamond with 20 at. % C were mechanically alloyed (MA) and consolidated via hot extrusion or spark plasma sintering (SPS). The hardness evolutions as well as the structural characterization of as-milled nanocomposite powders and consolidated samples are reported. Density measurements indicate that the consolidation outcome varies significantly with the process used. Transmission electron microscopy (TEM) inspection of the extrusion consolidated sample revealed bonding at the interface between copper and nDiamond particles. The nDiamond size distribution was determined from TEM observations. The results obtained are discussed in terms of consolidation routes.

W-Diamond/Cu-Diamond nanostructured composites for fusion devices

A novel material design for nuclear fusion reactors is proposed based on Cu-Diamond and W-Diamond nanocomposites. The proposed design involves the production of W/W-Diamond/CuDiamond/Cu functionally graded material. W, W-Diamond, Cu-Diamond and Cu nanostructured composite powders were produced independently by mechanical alloying and subsequently consolidated/welded through spark plasma sintering. Modulation of processing parameters allowed controlling the extent of unfavourable conversion of Diamond into tungsten carbide, as well as to overcome the Diamond intrinsically difficult bonding to copper. Microstructural features and microhardness of the as-produced materials are presented.

Helium load on W-O coatings grown by pulsed laser deposition

Abstract W-O deposits with complex morphologies and significant He contents will growth on the surface of plasma facing components exposed to He discharges. To mimic the re/co-deposition process, W-O coatings were loaded with He by implanting He+ ions on W films grown by pulsed laser deposition (PLD). The use of appropriate PLD experimental parameters such as pressurised Ar or He background atmospheres induces the deposition of porous or compact W structures enhancing afterwards the gathering of different amounts of O under exposure to atmospheric air. After multiple ion implantation stages using 150 keV, 100 keV and 50 keV incident He+ ion beams with a total fluence of 5 × 1017 ion/cm2, significant amounts of He were identified in porous coatings by Rutherford backscattering (RBS). Time-of-flight elastic recoil detection (ToF-ERDA) measurements showed that most of the implanted He was already released from the porous coatings five month after implantation, while for the case of compact layers the He content remains significant at deeper layers and smoothly decrease towards the surface, as result of a different morphology and nanostructure. The proposed method involving PLD and ion implantation seems adequate to produce W-O films load by He that may be used as reference samples for fusion investigations.

Raman microscopy investigation of beryllium materials

We report for the first time on the ability of Raman microscopy to give information on the structure and composition of Be related samples mimicking plasma facing materials that will be found in ITER. For that purpose, we investigate two types of material. First: Be, W, Be1W9, and Be5W5 deposits containing a few percents of D or N, and second: a Mo mirror exposed to plasma in the main JET chamber (in the framework of the first mirror test in JET with ITER-like wall). We performed atomic quantifications using ion beam analysis for the first samples. We also did atomic force microscopy. We found defect induced Raman bands in Be, Be1W9, and Be5W5 deposits. Molybdenum oxide has been identified showing an enhancement due to a resonance effect in the UV domain.

Studies on deuterium retention in W-Ta based materials

** *IST/ITN, Instituto Superior Tecnico, Instituto Tecnologico e Nuclear, Universidade Tecnica de Lisboa, Estrada Nacional 10, P-2686-953 Sacavem, Portugal **Associacao Euratom/IST, Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, Av. Rovisco Pais, 1049-001, Lisboa, Portugal ***LNEG, Laboratorio Nacional de Energia e Geologia, Estrada do Paco do Lumiar, 1649-038 Lisboa, Portugal ****ICEMS, Instituto Superior Tecnico, Universidade de Lisboa, Av. Rovisco Pais, 1049-001 Lisboa, Portugal *****National Institute of Advanced Industrial Science and Technology (AIST), 1-2-1 Namiki, Tsukuba, Ibaraki 305-8564, Japan ******CENIMAT-I3N, Departamento de Ciencia dos Materiais, Faculdade de Ciencias e Tecnologia, FCT, Universidade Nova de Lisboa, 2829-516 Caparica, Portugal The high melting point, high sputtering threshold and low tritium inventory rendered W as a potentially suitable material in fusion devices [1-4]. The major problem associated with presently available tungsten grades as structural materials is its brittleness at lower temperatures. This is further worsened by irradiation embrittlement. A strategy for ductility improvement is producing a composite, with the brittle W matrix being reinforced by short fibres of tantalum [5]. As Ta is more ductile than W it can therefore divert or stop cracks propagating in the W matrix. In the present research Ta short fibres and powder were used as reinforcement component for W [6] by alloying Ta short fibres or powder in a W powder matrix. The composites were subsequently irradiated with deuterium to assess the retention of this hydrogenic species in the materials. The irradiated composites, with Ta contents of 10 or 20 at%, were produced from pure elemental powders (W-Ta powder composites), and pure W powder and Ta fibre (W-Ta fibre composites) with 100 μm in diameter by low energy ball milling in argon atmosphere. These materials were consolidated via spark plasma sintering (SPS) in the temperature 1200 to 1600 oC range. Pure W and Ta plates (controls) and W-Ta composites were irradiated with He

Effects of hydrogen permeation on W, Mo and Cu Langmuir probes at ISTTOK

The microstructures of tungsten, molybdenum and copper wires used as Langmuir probes at ISTTOK edge plasma have been investigated. The probes cross-sections evidenced extensive grain growth, intergranular bubbles and increased hardness at the plasma exposed regions. Internal surfaces of large bubbles exhibited slip bands resulting from plastic deformation induced by high H2 pressure. Elastic recoil detection analysis was used to measure H concentration profiles. The present results suggest that H2 bubble formation in first wall components under long-term high thermal loads should be closely monitored in nuclear fusion devices. Strategies for H damage mitigation are proposed and discussed.

Microstructure characterization of ODS-RAFM steels

Results of the microstructural characterization of four different RAFM ODS Eurofer 97 batches are presented and discussed. Analyses and observations were performed by nuclear microprobe and scanning and transmission electron microscopy. X-ray elemental distribution maps obtained with proton beam scans showed homogeneous composition within the proton beam spatial resolution and, in particular, pointed to a uniform distribution of ODS (yttria) nanoparticles in the Eurofer 97 matrix. This was confirmed by transmission electron microscopy. Scanning electron microscopy coupled with energy dispersive spectroscopy made evident the presence of chromium carbide precipitation. Precipitates occurred preferentially along grain boundaries (GB) in three of the batches and presented a discrete distribution in the other, as a result of different thermo-mechanical routes. Additional electron backscattered diffraction experiments revealed the crystalline textures in the ferritic polycrystalline structure of the ODS steel samples.