E. Tejado

Bio: E. Tejado is an academic researcher from Technical University of Madrid. The author has contributed to research in topics: Tungsten & Flexural strength. The author has an hindex of 12, co-authored 26 publications receiving 424 citations. Previous affiliations of E. Tejado include Spanish National Research Council.

Evolution of mechanical performance with temperature of W/Cu and W/CuCrZr composites for fusion heat sink applications

Abstract: Power exhaust and materials lifetime have been identified as key challenges for next-generation fusion devices. A water-cooled monoblock divertor, consisting of tungsten as the plasma facing material and copper-based composites as the heat sink, has been proposed as the baseline material. However, there is a large mismatch in the coefficient of thermal expansion and the elastic modulus between these metals, which requires the development of new materials. The goal of this study is the mechanical and microstructural characterization of two composites materials, W-30 wt%Cu and W-30 wt%CuCrZr, produced using liquid infiltration in an open porous tungsten preform. To reproduce the most adverse in-service conditions, materials were characterized under high vacuum atmosphere up to 800 °C. From the measured mechanical properties, a remarkable temperature dependence can be assessed, since it is noticeable that for both materials, tensile and flexural strength equally decrease as the measurement temperature increases. However, the fracture behavior of W-30CuCrZr is on average 25% higher than that of W-30Cu (21 MPa m1/2 versus 17 MPa m1/2 at RT) although this gap is narrower at higher temperatures; at 800 °C, the contribution of the Cu-based phase is quite low, thus fracture is primary controlled by the W initial skeleton. From these values, it can be inferred that the yield strength and fracture toughness of W-30CuCrZr composite are superior, whilst it presented lower elastic modulus and rupture strain values than the W-30Cu composite. As a result, the metal matrix composites presented in this article could effectively dissipate heat while overcoming the thermal stresses produced during operation, since a decent thermomechanical performance was observed at relevant reactor temperatures. This is of vital importance to enhance the performance, life cycle, and reliability of the component.

Manufacturing and testing of self-passivating tungsten alloys of different composition

Abstract: Abstract Self-passivating tungsten based alloys for the first wall armour of future fusion reactors are expected to provide a major safety advantage compared to pure tungsten in case of a loss of coolant accident with simultaneous air ingress, due to the formation of a stable protective scale at high temperatures in presence of oxygen which prevents the formation of volatile and radioactive WO 3 . Bulk W-15Cr, W-10Cr-2Ti and W-12Cr-0.5Y alloys were manufactured by mechanical alloying followed by can encapsulation and HIP. This route resulted in fully dense materials with nano-structured grains. The ability of Ti and especially of Y to inhibit grain growth was observed in the W-10Cr-2Ti and W-12Cr-0.5Y alloys. Besides, Y formed Y-rich oxide nano-precipitates at the grain boundaries, and is thus expected to improve the mechanical behaviour of the Y-containing alloy. Isothermal oxidation tests at 800 oC (1073 K) and oxidation tests under accident-like conditions revealed that the W-12Cr-0.5Y alloy exhibits the best oxidation behaviour of all alloys, especially in the accident-like scenario. Preliminary HHF tests performed at GLADIS indicated that the W-10Cr-2Ti alloy is able to withstand power densities of 2 MW/m 2 without significant damage of the bulk structure. Thermo-shock tests at JUDITH-1 to simulate mitigated disruptions resulted in chipping of part of the surface of the as-HIPed W-10Cr-2Ti alloy. An additional thermal treatment at 1600 °C (1873 K) improves the thermo-shock resistance of the W-10Cr-2Ti alloy since only crack formation is observed.

Self-passivating tungsten alloys of the system W-Cr-Y for high temperature applications

Abstract: Self-passivating tungsten based alloys for the first wall armor of future fusion reactors are expected to provide a major safety advantage compared to pure tungsten in case of a loss-of-coolant accident with simultaneous air ingress, due to the formation of a stable protective scale at high temperatures in presence of oxygen which prevents the formation of volatile and radioactive WO3. This work analyses the oxidation and thermal shock resistance of W-Cr-Y alloys obtained by mechanical alloying followed by HIPing. Alloys with different Cr and Y contents are produced in fully dense form with nanocrystalline or ultrafine-grained microstructure and a dispersion of Y-rich oxide nanoparticles located mainly at the grain boundaries. Isothermal oxidation experiments confirm an excellent oxidation resistance due to the formation of protective oxide scales at the very surface. These layers mainly consist of Cr2O3 and mixed Y-W and Cr-W oxides. The superior oxidation resistance of these alloys is confirmed by tests simulating accident-like conditions. The thermal conductivity of these alloys at 600–1000 °C is 2–3 times higher than standard Ni-base superalloys like Inconel-718. The material also exhibits outstanding thermal shock resistance: 1000 pulses of 0.19 GW/m2 power density and 1 ms duration at 400 °C base temperature resulted in no damage, while an increased power density of 0.38 GW/m2 resulted in the formation of a crack-network and slight surface roughening. An additional thermal treatment at 1550 °C improves slightly the oxidation resistance and significantly the thermal shock resistance of the alloy.

Development of advanced high heat flux and plasma-facing materials

Abstract: Plasma-facing materials and components in a fusion reactor are the interface between the plasma and the material part. The operational conditions in this environment are probably the most challenging parameters for any material: high power loads and large particle and neutron fluxes are simultaneously impinging at their surfaces. To realize fusion in a tokamak or stellarator reactor, given the proven geometries and technological solutions, requires an improvement of the thermo-mechanical capabilities of currently available materials. In its first part this article describes the requirements and needs for new, advanced materials for the plasma-facing components. Starting points are capabilities and limitations of tungsten-based alloys and structurally stabilized materials. Furthermore, material requirements from the fusion-specific loading scenarios of a divertor in a water-cooled configuration are described, defining directions for the material development. Finally, safety requirements for a fusion reactor with its specific accident scenarios and their potential environmental impact lead to the definition of inherently passive materials, avoiding release of radioactive material through intrinsic material properties. The second part of this article demonstrates current material development lines answering the fusion-specific requirements for high heat flux materials. New composite materials, in particular fiber-reinforced and laminated structures, as well as mechanically alloyed tungsten materials, allow the extension of the thermo-mechanical operation space towards regions of extreme steady-state and transient loads. Self-passivating