The present work proposes a brazing procedure to join a self-passivating tungsten alloy (W-10Cr-0.5Y) with Eurofer. The results indicated the achievement of high continuity W-Eurofer joints using 80Cu-20Ti filler material, which involves operative brazeability. The resulting microstructure of the braze changes considerably compared to the brazing attempts with pure tungsten, which is associated to the reactive character of chromium. A high interaction between molten filler and W base material close to the braze was detected, with preferential grain boundary penetration of Cu-Ti into W alloy. It gave rise to a loss of hardness at the W base material near the joint. Regarding to the strength of the joints, shear strength of similar to 90 MPa was obtained.

Compared to pure tungsten, self-passivating tungsten based alloys for the first wall armor of future fusion reactors shall provide a major safety advantage 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 preventing the formation of volatile and radioactive WO3. Recently developed W-Cr-Y alloys produced by mechanical alloying and hot isostatic pressing (HIP) exhibit a strong reduction of oxidation rate compared to pure W and high mechanical strength. A heat treatment after HIP at 1555°C results in a one-phase material with a high thermal shock resistance. Nevertheless, the microstructure is metastable and its thermal stability under operational conditions has to be assessed. In this work results of thermal stability tests on heat treated W-10Cr-0.5Y alloy subjected to temperatures of 650, 700, 800 and 1000°C for times ranging from 50 to 1000h are presented. After 1000h at 650°C and 100h at 700°C no visible change of the microstructure is detected. After 100h at 1000°C a complete decomposition takes place with the formation of a uniform, fine-scale mixture of W- and Cr-rich phases, typical for spinodal decomposition.

Porous silicon carbide (SiC) is a promising ceramic for high-temperature applications due to its unique combination of properties. In the present work, a fabrication route for porous SiC is presented using graphite spherical powder as sacrificial phase to introduce porosity. By varying the initial amount of sacrificial phase, high-performance SiC materials with porosities in the range 30-50% were manufactured and characterized in terms of microstructure, density, thermal conductivity and flexural strength. The materials were fabricated by liquid phase sintering in presence of 2.5 wt.% Al2O3 and Y2O3 as sintering additives. The results indicate that the SiO2 present in the starting SiC powders interacts with the sintering additives forming liquid phases that promote densification and weight loss. Besides, an Al-Si liquid phase is formed at higher sintering temperatures, whose contribution to densification is inhibited in presence of graphite due to the formation of Al-rich carbides.

Six suppliers from several countries were asked to furnish parallelepiped 10 × 30 × 80 mm3 samples of commercial purity tungsten machined from rolled plates in view of selecting a provider of tungsten bricks for the target of the European neutron spallation source, ESS, under construction in Lund (Sweden). Sample surfaces were to be ground and free from visible defects or oxidation. The material should be rolled after sintering. A minimum room temperature tensile strength of 600 MPa was specified. The samples were submitted to different blind mechanical tests, measurement of physical properties and structural observations in order to assess their suitability for the application. We present here a summary of their main measured properties. The dispersion of results is noteworthy; the exercise allowed to sort-out technically eligible candidates for the application.

Flow channel inserts (FCIs) are the key elements in the high-temperature dual-coolant lead-lithium blanket, since in this concept the flowing PbLi reaches temperatures near 700 degrees C and FCIs should provide the necessary thermal and electrical insulations to assure a safe blanket performance. In this paper, the use of a SiC-sandwich material for FCIs consisting of a porous SiC core covered by a dense chemical vapor deposition-SiC layer is studied. A fabrication procedure for porous SiC is proposed and the resulting materials are characterized in terms of thermal and electrical conductivities (the latter before and after being subjected to ionizing radiation) and flexural strength. SiC materials with a wide range of porosities are produced; in addition, preliminary results using an alternative route based on the gel-casting technique are also presented, including the fabrication of hollow samples to be part of future lab-scale FCI prototypes. Finally, to study the corrosion resistance of the material in hot PbLi, corrosion tests under static PbLi at 700 degrees C and under flowing PbLi at similar to 10 cm/s and 550 degrees C, with and without a 1.8-2T magnetic field, were performed to materials coated with a 200-400-mu m-thick dense SiC layer, obtaining promising results.

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 degrees 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/m(2) power density and 1 ms duration at 400 degrees C base temperature resulted in no damage, while an increased power density of 0.38 GW/m(2) resulted in the formation of a crack-network and slight surface roughening. An additional thermal treatment at 1550 degrees C improves slightly the oxidation resistance and significantly the thermal shock resistance of the alloy.

Porous SiC samples with different percentage of sintering additives have been manufactured by the so-called sacrificial template method at the Ceit-IK4 Technology Center (San Sebastian, Spain). Material stability under ionizing radiation and high temperature conditions considering electrical conductivity and microstructure has been evaluated at Ciemat (Madrid, Spain). Electrical conductivity was measured as a function of temperature before and after irradiation with 1.8 MeV electrons up to 23 MGy (similar to 10(-6) dpa), and radiation induced conductivity (RIC) was also examined during irradiation at 550 degrees C for different dose rates (from 0.5 to 5 kGy/s). Electrical conductivity increase with irradiation dose was observed to occur. Posterior XRD analysis allowed interpret radiation induced modification of the electrical conductivity in terms of changes in the SiC crystalline structure. Although neither RIC nor electrical degradation is seen to be an issue when data is extrapolated to future fusion devices radiation levels, the observed phase transformation at relatively low ionizing dose might affect structural stability of SiC-based materials.

Tungsten is currently deemed as a promising plasma-facing material (PFM) for the future power plant DEMO. In the case of an accident, air can get into contact with PFMs during the air ingress. The temperature of PFMs can rise up to 1200 °C due to nuclear decay heat in the case of damaged coolant supply. Heated neutron-activated tungsten forms a volatile radioactive oxide which can be mobilized into the atmosphere. New self-passivating 'smart' alloys can adjust their properties to the environment. During plasma operation the preferential sputtering of lighter alloying elements will leave an almost pure tungsten surface facing the plasma. During an accident the alloying elements in the bulk are forming oxides thus protecting tungsten from mobilization. Good plasma performance and the suppression of oxidation are required for smart alloys. Bulk tungsten (W)¿chroimum (Cr)¿titanium (Ti) alloys were exposed together with pure tungsten (W) samples to the steady-state deuterium plasma under identical conditions in the linear plasma device PSI 2. The temperature of the samples was ~576 °C¿715 °C, the energy of impinging ions was 210¿eV matching well the conditions expected at the first wall of DEMO. Weight loss measurements demonstrated similar mass decrease of smart alloys and pure tungsten samples. The oxidation of exposed samples has proven no effect of plasma exposure on the oxidation resistance. The W¿Cr¿Ti alloy demonstrated advantageous 3-fold lower mass gain due to oxidation t

Material issues pose a significant challenge for future fusion reactors like DEMO: When using materials in a fusion environment a highly integrated approach is required. Damage resilience, power exhaust, as well as oxidation resistance during accidental air ingress are driving issues when deciding for new materials. Neutron induced effects, e.g. transmutation adding to embrittlement is crucial to material performance. Here advanced materials, e.g. W-f/W or W/Cu, W-f/Cu composites allow the step towards a fusion reactor. Recent developments in the area of multi-fibre powder-metallurgical Wf/W mark a possible path towards a component based on standard tungsten production technologies. Field assisted sintering technology is used as production route to achieve 94% dense materials. Initial mechanical tests and micro-structural analyses show potential for pseudo-ductile behavior of materials with a reasonable (30%) fibre fraction. In the as-fabricated condition samples showed step-wise cracking while the material is still able to bear rising load, the typical pseudo-ductile behavior of a composite. Yttria is used as the interface material in order to allow the energy dissipation mechanisms to become active. W-f/W as plasma facing material contributes here to advanced material strength and crack resilience even with a brittle matrix embrittlement, while W/Cu, W-f/Cu composites at the coolant level allow for higher strength at elevated cooling temperatures. In addition to the use of pure tungsten it is demonstrated that tungsten based self-passivating alloys can also be used in the composite approach. (C) 2017 The Authors. Published by Elsevier B.V.

The use of self-passivating tungsten alloys for the first wall armor of future fusion reactors is advantageous concerning safety issues in comparison with pure tungsten. Bulk W-10Cr-0.5Y alloy manufactured by mechanical alloying followed by HIP resulted in a fully dense material with grain size around 100 nm and a dispersion of Y-rich oxide nanoparticles located at the grain boundaries. An improvement in flexural strength and fracture toughness was observed with respect to previous works. Oxidation tests under isothermal and accident-like conditions revealed a very promising oxidation behavior for the W-10Cr-0.5Y alloy. Thermo-shock tests at JUDITH-1 to simulate ELM-like loads resulted in a crack network at the surface with roughness values lower than those of a pure W reference material. An additional thermal treatment at 1550 degrees C improves slightly the oxidation and significantly thermo-shock resistance of the alloy. (C) 2017 The Authors. Published by Elsevier B.V.

Flow Channel Inserts (FCIs) are key elements in a DCLL blanket concept for DEMO, since they provide the required thermal insulation between the He cooled structural steel and the hot liquid PbLi flowing at approximate to 700 degrees C, and the necessary electrical insulation to minimize MHD effects. In this work a SiC-based sandwich material is proposed for FCIs, consisting of a porous SiC core covered by a dense CVD-SiC layer. A method to produce the porous SiC core is presented, based on combining a starting mixture of SiC powder with a spherical carbonaceous sacrificial phase, which is removed after sintering by oxidation, in such a way that a microstructure of spherical pores is achieved. Following this technique, a porous SiC material with low thermal and electrical conductivities, but enough mechanical strength was produced. Samples were covered by a 200 mu m thick CVD-SiC coating to form a SiC-sandwich material. Finally, corrosion tests under static PbLi were performed, showing that such a dense layer offers a reliable protection against static PbLi corrosion. (C) 2017 Elsevier B.V. All rights reserved.

An ultra-fine grained self-passivating tungsten alloy (W88-Cr10-Ti2 in wt.%) has been implanted with iodine ions to average doses of 0.7 and 7 dpa, as well as with helium ions to an average concentration of 650 appm. Pile-up corrected Berkovich nanoindentation reveals significant irradiation hardening, with a maximum hardening of 1.9 GPa (17.5%) observed. The brittle fracture strength of the material in all implantation conditions was measured through un-notched cantilever bending at the microscopic scale. All cantilever beams failed catastrophically in an intergranular fashion. A statistically confirmed small decrease in strength is observed after low dose implantation (-6%), whilst the high dose implantation results in a significant increase in fracture strength (+9%), further increased by additional helium implantation (+16%). The use of iodine ions as the implantation ion type is justified through a comparison of the hardening behaviour of pure tungsten under tungsten and iodine implantation.

Tungsten is the major candidate material for the armour of plasma facing components in future fusion devices. To overcome the intrinsic brittleness of tungsten, which strongly limits its operational window, a W-fibre enhanced W-composite material (W-f/W) has been developed incorporating extrinsic toughening mechanisms. Small WOW samples show a large increase in toughness. Recently, a large sample (50 mm x 50 mm x 3 mm) with more than 2000 long fibres has been successfully produced allowing further mechanical and thermal testing. It could be shown that even in a fully embrittled state, toughening mechanisms as crack bridging by intact fibres, as well as the energy dissipation by fibre-matrix interface debonding and crack deflection are still effective. A potential problem with the use of pure Win a fusion reactor is the formation of radioactive and highly volatile WO3 compounds and their potential release under accidental conditions. It has been shown that the oxidation of W can be strongly suppressed by alloying with elements forming stable oxides. WCr10Ti(2) alloy has been produced on a technical scale and has been successfully tested in the high heat flux test facility GLADIS. Recently, W-Cr-Y alloys have been produced on a lab-scale. They seem to have even improved properties compared to the previously investigated W alloys

Tungsten is the major candidate material for the armour of plasma facing components in future fusion devices. To overcome the intrinsic brittleness of tungsten, which strongly limits its operational window, a W-fibre enhanced W-composite material (W-f/W) has been developed incorporating extrinsic toughening mechanisms. Small WOW samples show a large increase in toughness. Recently, a large sample (50 mm x 50 mm x 3 mm) with more than 2000 long fibres has been successfully produced allowing further mechanical and thermal testing. It could be shown that even in a fully embrittled state, toughening mechanisms as crack bridging by intact fibres, as well as the energy dissipation by fibre-matrix interface debonding and crack deflection are still effective. A potential problem with the use of pure Win a fusion reactor is the formation of radioactive and highly volatile WO3 compounds and their potential release under accidental conditions. It has been shown that the oxidation of W can be strongly suppressed by alloying with elements forming stable oxides. WCr1OTi(2) alloy has been produced on a technical scale and has been successfully tested in the high heat flux test facility GLADIS. Recently, W-Cr-Y alloys have been produced on a lab-scale. They seem to have even improved properties compared to the previously investigated W alloys. (C) 2016 EURATOM. Published by Elsevier B.V. All rights reserved.

The brittle fracture strength of a self-passivating W-Cr10-Ti2 alloy (in wt.%) was measured through un-notched cantilever bending at the microscopic scale. The material behaved purely elastic and fractured catastrophically in an unstable fashion. An average nominal strength of 5.9 GPa was measured. The scatter in strength was shown to be significantly higher than the sum of all random errors indicating an inherent variability of the material¿s strength. The measurements from 28 tests followed a Weibull distribution with a modulus of m = 12. Results from a size effect study at the microscopic scale were successfully predicted through Weibull scaling. Extrapolation into the macroscopic range overestimated the measured three-point bend strength, which is likely due to the presence of large-scale heterogeneities. The test technique sampled a material thickness of only several micrometres and is hence suitable for future ion irradiation studies.

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 WO3. 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 degrees C (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 degrees C (1873 K) improves the thermo-shock resistance of the W-10Cr-2Ti alloy since only crack formation is observed. (C) 2016 The Authors. Published by Elsevier Ltd.

Self-passivating tungsten based alloys will provide a major safety advantage compared to pure tungsten when used as first wall armor of future fusion reactors, due to the formation of a protective oxide layer which prevents the formation of volatile and radioactive WO3 in case of a loss of coolant accident with simultaneous air ingress. Bulk WCr10Ti2 alloys were manufactured by two different powder metallurgical routes: (1) mechanical alloying (MA) followed by hot isostatic pressing (HIP) of metallic capsules, and (2) MA, compaction, pressureless sintering in H-2 and subsequent HIPing without encapsulation. Both routes resulted in fully dense materials with homogeneous microstructure and grain sizes of 300 nm and 1 mu m, respectively. The content of impurities remained unchanged after HIP, but it increased after sintering due to binder residue. It was not possible to produce large samples by route (2) due to difficulties in the uniaxial compaction stage. Flexural strength and fracture toughness measured on samples produced by route (1) revealed a ductile-to-brittle-transition temperature (DBTT) of about 950 degrees C. The strength increased from room temperature to 800 degrees C, decreasing significantly in the plastic region. An increase of fracture toughness is observed around the DBTT.

The conventional PM ODS Ferritic Steel (FS) processing route includes gas atomisation of steel powder and its mechanical alloying (MA) with Y2O3 powder particles to dissolve yttrium and form, during consolidation, a dispersion of oxide nanoparticles (Y-Ti-O) in a nanostructured matrix. This work presents an alternative route to produce ODS steels avoiding MA: STARS (Surface Treatment of gas Atomized powder followed by Reactive Synthesis). STARS FS powders with composition Fe-14Cr-2W-0.3Ti-0.23Y, already containing the nanoparticles precursors, were gas-atomized. Oxygen, Y and Ti contents were tailored to the required values to form Y-Ti-O nanoparticles during processing. Powders were HIPped at 900, 1220 and 1300 degrees C. Specimens HIPped at 900 and 1220 degrees C were heat treated (HT) at temperatures ranging from 1200 to 1320 degrees C. The microstructural evolution with HIP and HT temperatures, including characterisation of nanoparticles and feasibility of achieving complete dissolution of prior particle boundaries (PPBs) were assessed.

The Dual Coolant Led Lithium (DCLL) blanket is one of the concepts being investigated as candidate for DEMO, due to the high thermal efficiency provided by the flowing PbLi self-cooled breeder at approximate to 700 degrees C in the high temperature design. Key elements are the Flow Channel Inserts (FCIs) serving as electrical and thermal insulators to mitigate MHD effects and to keep the He-cooled steel walls below its maximum allowable temperature due to corrosion. A material based on sandwiching porous SiC between dense SiC layers is proposed for FCIs. In this work results of theoretical calculations and an FEM model are presented to determine the optimum thickness of both porous core and outer dense layers to assure the required thermal insulation across the FCI with minimum thermal stresses, considering achievable properties for the porous SiC material and its fabrication possibilities. It is concluded that the porous core thickness must be at least 5 mm if a porous SiC with thermal conductivity around 7 W/mK is used; a dense coating of approximate to 200 mu m is considered as optimum regarding the thermal stresses present in the FCI.

Several mock-ups, each of them consisting of six rectangular channels with dimensions according to the EU Test Blanket Modules (TBMs) specifications, were manufactured by selective laser melting (SLM) technology using P91, a ferritic-martensitic 9%Cr-1%Mo-V steel with a metallurgical behavior similar to EUROFER, the reference structural material for DEMO blanket concepts. SLM parameters led to an as-built density of 99.35% Theoretical Density (TD) that increased up to 99.74% after hot isostatic pressing (HIP). Dimensional control showed that the differences between the original design and the component are below 100 pm. By the appropriate selection of normalization and tempering parameters it was possible to obtain a material fulfilling P91 specification. The microstructure was investigated after SLM, HIP and normalizing and tempering treatments. In all cases, it consisted of thin martensitic laths. Subsize tensile samples were extracted from the mock-ups to measure the mechanical tensile properties after each step of the manufacturing process. The effect of thermal treatments on hardness was also evaluated.

Nanostructured Oxide Dispersion Strengthened Reduced Activation Ferritic Stainless Steels (ODS RAF) are promising structural materials for fusion reactors, due to their ultrafine microstructure and the presence of a dispersion of Y-Ti-O nanoclusters that provide excellent creep strength at high temperatures (up to 750 °C). The traditional powder metallurgical route to produce these steels is based on Gas Atomization (GA) + Mechanical Alloying (MA) + HIP + ThermoMechanical Treatments (TMTs). Recently, alternative methods have arisen to avoid the MA step. In line with this new approach, ferritic stainless steel powders were produced by gas atomization and HIPped, after adjusting their oxygen, Y and Ti contents to form Y¿Ti¿O nanoclusters during subsequent heat treatments. The microstructure of as-HIPped steels mainly consists of ferrite grains, Y-Ti precipitates, carbides and oxides on Prior Particle Boundaries (PPBs). Post-HIP heat treatments performed at high temperatures (1270 and 1300 °C) evaluated the feasibility of achieving a complete dissolution of the oxides on PPBs and a precipitation of ultrafine Ti- and Y-rich oxides in the Fe14Cr2W matrix. FEG-SEM with extensive EDS analysis was used to characterize the microstructure of the atomized powders and the ODS-RAF specimens after HIP consolidation and post-HIP heat treatments. A deeper characterization of atomized powder was carried out by TEM.

Self-passivating tungsten based alloys are expected to provide a major safety advantage compared to pure tungsten when used as first wall armour of future fusion reactors, due to the formation of a protective oxide scale, preventing the formation of volatile and radioactive WO3 in case of a loss of coolant accident with simultaneous air ingress. In this work results of isothermal oxidations tests at 800 and 1000 degrees C on bulk alloy WCr12Ti2.5 performed by thermogravimetric analysis (TGA) and by exposure to flowing air in a furnace are presented. In both cases a thin, dense Cr2O3 layer is found at the outer surface, below which a Cr2WO6 scale and Ti2CrO5 layers alternating with WO3 are formed. The Cr2O3, Cr2WO6 and Ti2CrO5 scales act as protective barriers against fast inward O2- diffusion. The oxidation kinetics seems to be linear for the furnace exposure tests while for the TGA tests at 800 degrees C the kinetics is first parabolic, transforming into linear after an initial phase. The linear oxidation rates are 2-3 orders of magnitude lower than for pure W.

The structure and crystallographic texture of zinc strips (Zn-Cu-Ti alloy) produced by the continuous horizontal twin roll strip casting method has been characterized. In longitudinal sections normal to the transverse direction, the strips display an approximately symmetrical chevron patterned structure of columnar grains inclined about 30 degrees from the rolling direction. In association with such structure, the macroscopic texture is mainly < 1 (1) over bar 00 > 'normal' (not cyclic) fibre texture tilted approximately +/-30 degrees around the transverse direction plus a similarly tilted weak < 0001 > fibre texture. A thin layer of small equiaxed grains with a strong (0001) basal texture is present at the free surfaces. The observed structure/texture combination agrees quite well with the expected macrostructure of solidification of the alloy in the twin roll casting process.

Thermally and electrically insulating porous SiC ceramics are attractive candidates for Flow Channel Inserts (FCI) in dual-coolant blanket concepts thanks to its relatively inexpensive manufacturing route. To prevent tritium permeation and corrosion by Pb-15.7 a dense coating has to be applied on the porous SiC. Despite not having structural function, FCI must exhibit sufficient mechanical strength to withstand strong thermal gradients and thermo-electrical stresses during operation. This work summarizes the results on the development of coated porous SiC for FCI. Porous SiC was obtained following the sacrificial template technique, using Al2O3 and Y2O3 as sintering additives and a carbonaceous phase as pore former. Sintering was performed in inert gas at 1850-1950 degrees C during 15 min to 3 h, followed by oxidation at 650 degrees C to eliminate the carbonaceous phase. The most promising bulk materials were coated with a similar to 30 mu m thick dense SiC by CVD. Results on porosity, bending tests, thermal and electrical conductivity are presented. The microstructure of the coating, its adhesion to the porous SiC and its corrosion behavior under Pb-17.5Li are also shown.

The long-term objective of the European Fusion Development Agreement (EFDA) fusion materials programme is to develop structural and armor materials in combination with the necessary production and fabrication technologies for reactor concepts beyond the International Thermonuclear Experimental Reactor. The programmatic roadmap is structured into four engineering research lines which comprise fabrication process development, structural material development, armor material optimization, and irradiation performance testing, which are complemented by a fundamental research programme on "Materials Science and Modeling." This paper presents the current research status of the EFDA experimental and testing investigations, and gives a detailed overview of the latest results on materials research, fabrication, joining, high heat flux testing, plasticity studies, modeling, and validation experiments.

Self-passivating tungsten based alloys are expected to provide a major safety advantage compared to pure tungsten, presently the main candidate material for first wall armour of future fusion reactors. In case of a loss of coolant accident with simultaneous air ingress, a protective oxide scale will be formed on the surface of W avoiding the formation of volatile and radioactive WO3. Bulk WCr12Ti2.5 alloys were manufactured by mechanical alloying (MA) and hot isostatic pressing (HIP), and their properties compared to bulk WCr10Si10 alloys from previous work. The MA parameters were adjusted to obtain the best balance between lowest possible amount of contaminants and effective alloying of the elemental powders. After HIP, a density >99% is achieved for the WCr12Ti2.5 alloy and a very fine and homogeneous microstructure with grains in the submicron range is obtained. Unlike the WCr10Si10 material, no intergranular ODS phase inhibiting grain growth was detected. The thermal and mechanical properties of the WCr10Si10 material are dominated by the silicide (W,Cr)(5)Si-3; it shows a sharp ductile-to brittle transition in the range 1273-1323 K. The thermal conductivity of the WCr12Ti2.5 alloy is close to 50 W/mK in the temperature range of operation; it exhibits significantly higher strength and lower DBTT - around 1170 K - than the WCr10Si10 material.

Tungsten materials are candidates for plasma-facing components for the International Thermonuclear Experimental Reactor and the DEMOnstration power plant because of their superior thermophysical properties. Because these materials are not common structural materials like steels, knowledge and strategies to improve the properties are still under development. These strategies discussed here, include new alloying approaches and microstructural stabilization by oxide dispersion strengthened as well as TiC stabilized tungsten based materials. The fracture behavior is improved by using tungsten laminated and tungsten wire reinforced materials. Material development is accompanied by neutron irradiation campaigns. Self-passivation, which is essential in case of loss-of-coolant accidents for plasma facing materials, can be achieved by certain amounts of chromium and titanium. Furthermore, modeling and computer simulation on the influence of alloying elements and heat loading and helium bombardment will be presented.

The erosion yield by deuterium impact was determined for various doped carbon-based materials. Ion beam bombardment with 30 and 200 eV at elevated temperatures (600-850 K) and low temperature plasma exposure with 30 eV ion energy (similar to 7 x 10(20) ions/m(2)s) and about 170 times higher thermal atomic deuterium flux at 300 K and 630 K were performed. The total yield of fine-grain graphites doped with 4 at.% Ti and Zr is reduced by a factor of 4 for 30 and 200 eV D impact at elevated temperatures at D fluences above 10(24) m(-2) compared to undoped graphite. Extensive carbide particle loss can be excluded up to fluences of similar to 10(25) m(-2).

SiC is the primary candidate for the flow channel inserts in dual-coolant blanket concepts. Porous SiC ceramics are attractive candidates for this non-structural application, since they can satisfy the required properties through a low cost manufacturing route, compared to SiC(f)/SiC. This work shows first results of the manufacturing of porous SiC ceramics prepared with different amounts of Y(2)O(3) and Al(2)O(3) as sintering additives. C powders were used as pore-formers by their burnout during oxidation after sintering. Comparison of microstructure, porosity, flexural strength, thermal and electrical conductivity and corrosion under Pb-15.7Li of porous SiC without and with sintering additives is presented. The addition of 2.5 wt.% of Y(2)O(3) and Al(2)O(3) improves the mechanical properties, and reduces the thermal and electrical conductivity down to reasonable values. Preliminary corrosion tests under Pb-15.7 Li at 500 degrees C show that the absence of a dense coating on porous SiC leads to poor corrosion behavior. (C) 2011 Elsevier B.V. All rights reserved.

Self-passivating tungsten-based alloys may provide a major safety advantage in comparison with pure tungsten, which is presently the main candidate material for the plasma-facing protection of future fusion power reactors. WCrSi alloys were manufactured by mechanical alloying (MA) and HIP at 1300 degrees C and 200 MPa for 1 h. Different MA conditions were investigated to obtain powders with lowest possible amount of contaminants and small and homogeneous particle and crystallite size. Milling in WC vials under Ar without process control agent provided best results. After HIP densities close to 100% were obtained. First oxidation tests on preliminary alloys showed self-passivating behavior with rates comparable to WCrSi thin films at 800 degrees C but worse performance at 1000 degrees C. In all cases a Cr(2)WO(6) protective layer is formed at the surface.

All the recent DEMO design studies for helium cooled divertors utilize tungsten materials and alloys, mainly due to their high temperature strength, good thermal conductivity, low erosion, and comparably low activation under neutron irradiation. The long-term objective of the EFDA fusion materials programme is to develop structural as well as armor materials in combination with the necessary production and fabrication technologies for future divertor concepts. The programmatic roadmap is structured into four engineering research lines which comprise fabrication process development, structural material development, armor material optimization, and irradiation performance testing, which are complemented by a fundamental research programme on "Materials Science and Modeling". This paper presents the current research status of the EFDA experimental and testing investigations, and gives a detailed overview of the latest results on fabrication, joining, high heat flux testing, plasticity, modeling, and validation experiments.

Self-passivating tungsten-based alloys are expected to provide a major safety advantage compared to pure tungsten, which is at present the main candidate material for the first wall armour of future fusion reactors. WC10Si10 alloys were manufactured by mechanical alloying (MA) in a Planetary mill and subsequent hot isostatic pressing (HIP), achieving densities above 95%. Different MA conditions were studied. After MA under optimized conditions, a core with heterogeneous microstructure was found in larger powder particles, resulting in the presence of some large W grains after HIP. Nevertheless, the obtained microstructure is significantly refined compared to previous work. First MA trials were also performed on the Si-free system WCr12Ti2.5. In this case a very homogeneous structure inside the powder particles was obtained, and a majority ternary metastable bcc phase was found, indicating that almost complete alloying occurred. Therefore, a very fine and homogeneous microstructure can be expected after HIP in future work.

The porosity of undoped and Zr- and Ti-doped graphites has been determined using ultra small and small angle neutron scattering, nitrogen adsorption and helium pycnometry To differentiate between open and closed in the neutron measurements the contrast matching technique was employed It is shown that the combination of the three techniques is necessary for an accurate determination of the porosity and for the assessment of the structural modifications arising from the doping which results in the transformation of open porosity to closed