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Introduction: Oxide dispersion strengthened (ODS) steels have proven to be a viable candidate as a structural material in a fusion power plant, due to their enhanced mechanical properties and resilience under irradiation and at high temperatures. A high interest in the development of austenitic ODS (AODS) steel became noticeable, which is shown by the increased amount of publications in that field in the recent years. That can be related to the inherently better corrosion resistance and superior creep properties compared to its ferritic counterpart. However, one of the major drawbacks of AODS steels was the more challenging mechanical alloying (MA) process and a lower powder production yield, caused by a more ductile and adhesive powder. This disadvantage was tackled by the addition of a process control agent (PCA) during the MA, but it was yet to be shown how the addition of a PCA affects the microstructure of the AODS. Methods: AODS with a carbon-containing PCA was mechanical alloyed, hot-extruded, and subsequently annealed at 700, 900, and 1100 °C for 2, 750, 1000, 1250 and 1500 hours to investigation the influence of carbon on the formation and stability of precipitates as well as on the grain size in comparison to available literature data. Scanning and transmission electron microscopy and atom probe tomography methods were applied in this systematic study to identify the possible growth of nano-precipitates. An Arrhenius-equation was used to determine the activation energy of the growth of precipitates. Results and Discussion: A growth of precipitates is barely detectable at temperatures equal or lower than 900 °C. The grain size remains stable. However, an unexpected increase of the grain and precipitates sizes and a decreased activation energy was measured at temperatures of 1100 °C for all annealing times. Recently published results of similar AODS steels have shown a stability of the microstructure up to 1250 °C. Due to this contradiction, we concluded that the growth of precipitates and the reduction of the grain boundary pinning force was supported by the diffusion of carbon and the formation of carbides.

This paper addresses the EUROfusion DEMO materials research program and highlights the characterization of the in-vessel baseline materials EUROFER97, CuCrZr and tungsten as well as advanced structural and high heat flux materials developed for risk mitigation. In support of engineering design activities the focus primarily is to assemble a data base, to supply material property handbooks, material assessment reports, DEMO design criteria and material design limits for DEMO thermal, mechanical and environmental conditions. For advanced candidate material including (a) steels optimized towards lower or higher operational windows, (b) heat sink materials (copper alloys or composites) and (c) tungsten based plasma facing materials data representative for DEMO are presented. In view of preparing for future licensing and clarification of performance limits, as major step nine neutron irradiation campaigns were started recently. The campaigns include (i) baseline materials irradiated up to medium neutron dose levels, typical for DEMO 1stphase, to fill gaps in the materials property handbook and improve design limits (ii) advanced material irradiated at lower fluence for screening and down-selection and (iii) a huge temperature-dpa matrix at low fluence for material modelling verification and increased fundamental knowledge.

Tungsten (W) based materials for fusion reactor applications have been developed in Japan for the last decade to improve thermo-mechanical properties and tolerance with respect to neutron-irradiation effects. Potassium (K) doping for dispersion strengthening and alloying by rhenium (Re) for solid solute softening and strengthening were applied to W fabricated by powder metallurgy. Thereby modified W materials, e.g. K-doped W and K-doped W-3%Re, demonstrated a higher recrystallization temperature threshold, an increase in strength and ductility, and a reduction in ductile-to-brittle transition temperature (DBTT) compared to pure W in the non-irradiated state. Embrittlement caused by displacement damage and its amplification due to solid transmutation products is expected to be an intrinsic issue of W based materials under neutron irradiation environment inherent to future fusion reactors. In this paper, the effect of K-doping and alloying by Re on the neutron-irradiation-induced embrittlement of W was investigated at neutron fluence causing a considerable amount of solid transmutation products.

The effect of neutron irradiation on the microstructure of EUROFER-ODS alloy reinforced with an addition of 0.5% Y2O3 was analyzed by transmission electron microscopy. The irradiation of up to 16.2 dpa damage dose at 300°C and 400°C was performed in Petten High Flux Reactor. This study focuses on the imaging, identification, and quantitative analysis of interstitial loops and voids located on ODS particles. The nano-sized loops and point defects with b½〈111〉 Burgers vectors are preferentially located near structural defects such as the line dislocations or impurities in the material irradiated at 300°C. Homogeneously distributed loops having Burgers vectors of b½〈111〉 and b〈100〉 types were found in the material irradiated at 400°C. The obtained quantitative data on radiation-induced defects such as size, number density and relaxation volume were compared with EUROFER97 irradiated at the same conditions. It was found that the occurrence of radiation induced defects depends on the local number density and the size distribution of the ODS particles and not only on the irradiation damage and temperature.

By uncoupling the manufacturability from the design process, additive manufacturing of RAFM steels like EUROFER97 can open significant design freedom for components in fusion reactors, e.g., breeding blankets and divertor applications. As mechanical properties of RAFM steels are based on their specific microstructure and high number density of precipitates, good control of the microstructure of additively manufactured parts has to be ensured. As additive manufactured components are known to possess unique microstructures compared to EUROFER97 from standard technologies, the aim of this paper is to investigate additive manufactured EUROFER97 components and the influence of post-processing steps on their microstructure and in particular on the formation of precipitates. This paper covers the technological fabrication process of EUROFER97 by selective laser melting (SLM), including the production and chemical analysis of pre-alloyed EUROFER97 powder. In the initial state after fabrication, SLM-EUROFER97 components exhibit a bimodal, anisotropic microstructure with large ferritic grains instead of the desired homogeneous, fully martensitic microstructure. Further, neither M23C6 nor MX precipitates were detected in the as-built condition. A heat treatment including hot isostatic pressing, austenitization, quenching and tempering, allows achieving a fully martensitic, uniform microstructure as well as the formation of M23C6 and MX precipitates with similar size and volume density as in conventionally produced EUROFER97. In combination with the chemical analysis of the raw material, impurities stemming from the atomization process forming oxides were detected. However, due to their small size, those are not expected to influence the performance of SLM-processed EUROFER97. The comparison of the as-built condition and the condition after hot isostatic pressing and heat treatments reveals the necessity of appropriate post-processing to tailor the microstructure to one comparable to conventional EUROFER97.

This work addresses an overview of the recent progress, associated risks and development plans of structural materials for in-vessel components (IVCs) in DEMO plants. Reduced Activation Ferritic Martensitic Steels form the primary structural materials for most DEMOs IVCs and will be the focus of the paper. An overview of EU- and J-DEMO programs RAFM steels developments is presented. We highlight the present status in their development for use in DEMO plants, with a focus on the structural materials’ operational, and project-oriented, requirements. The development of materials property handbooks from high-quality data is illustrated. A process to validate these steels for operation in DEMO IVCs is summarised, revealing the pragmatic procedures ongoing, and limitations of this approach due to the synergistic operational effects on IVC materials; the use of in-situ or surveillance monitoring is outlined as a method to accommodate this limitation in synergistic effects and allow confidence in future DEMO operations. The development of advanced modifications to F82H and EUROFER are highlighted, with minor modifications leading to improved low and high-temperature operational design space being open for DEMO reactors.

Recent analysis result in significant impact on the design of the Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB). Especially the main demanding sub-component of the BB in terms of fabrication, the First Wall (FW) is highly affected. As a matter of fact, the relevance of the developments made for the industrial fabrication of the FW used e.g. in the EU HCPB Test Blanket Module (TBM) for ITER is limited: On one hand in terms of the overall dimensions, but also in terms of geometry and cooling channel configuration. Thus, a fabrication strategy for the FW was proposed as an option in 2018 based on Additive Manufacturing (AM) offering a solution to cover the new aspects. The dedicated fabrication routine is based on Cold Spray (CS) in alternation with machining. In terms of manufacturing of non-plasma facing complex shaped thin- or double wall BB structures (e.g. Fuel Pins proposed for the HCPB BB) the use of AM also provides benefits compared to conventional technologies. Therefore, the AM process of Selective Laser Melting (SLM) is also investigated in terms of applicability for BBs. This paper gives an overview about the advances in different AM options applied for nuclear fusion structural low activation steels. Possibilities for spin-offs to other technological fields are discussed and conclusions are drawn reflecting licensing aspects and technological limits.

Currently, for the EU DEMO, two Breeding Blankets (BBs) have been selected as potential candidates for the integration in the reactor. They are the Water Cooled Lithium Lead and the Helium Cooled Pebble Bed BB concepts. The two BB variants together with the associated ancillary systems drive the design of the overall plant. Therefore, a holistic investigation of integration issues derived by the BB and the installation of its ancillary systems has been performed. The issues related to the water activation due to the 16N and 17N isotopes and the impact on the primary heat transfer systems have been investigated providing guidelines and dedicated solution for the integration of safety devices as isolation valves. The tritium retention and the permeation rates through the blanket and its ancillary systems have been also assessed taking into account different operating points both for the BB and ancillaries and comparing, when possible, the releases with the operating and safety limits. Moreover, the issues related to the tritium start-up inventory as well as the uncertainties on the Tritium Breeding Ratio (TBR) due to the integration of the auxiliary systems within the Vacuum Vessel have been also studied. Finally, the impact of the BB concepts on the safety systems like the Vacuum Vessel Pressure Suppression System is described with a particular focus on the different measures that should be implemented according to the considered concept. All these aspects are then taken into account to drive future developments during the Concept Design Phase., peer-reviewed

In this work, we have performed neutron irradiation and sub-sequent microstructural and micro-hardness measurements on a series of tungsten grades. The neutron irradiation was performed in BR2 reactor up to 0.2 dpa at 600, 800 and 1200 °C. The selected irradiation temperatures correspond to a large range of the operation of tungsten in ITER as armour for the divertor. The applied fast neutron flux is representative of ITER irradiation conditions in terms of damage rate. The primary purpose of the study was to establish the relation between the initial microstructure, the resulting irradiation-induced microstructure and corresponding irradiation hardening. The microstructure was studied using transmission electron microscopy (TEM) and hardening was assessed by micro-hardness measurements. The obtained TEM results were used to predict the hardening employing the dispersed barrier model and strength coefficients established for the single crystal from the open literature. Based on the performed analysis, it appeared that the main contribution to the hardening at high irradiation temperature originates from the voids, while the dislocation loops provide a comparable contribution (30–50%) only under irradiation at 600 °C. Excellent agreement between the model prediction and experimentally measured hardness increase was obtained for the single crystal and ITER specification grade without any additional adjustment, while the hardening induced in the cold rolled plate was essentially overestimated by the model. This result suggests that the cold rolled tungsten plate exhibits an alternative mechanisms of the plastic deformation besides the conventional dislocation glide and multiplication, and this alternative mechanism is impacted by the irradiation less severely compared to the prediction coming from the dispersed barrier model.

The composition of new eigth 9% Cr reduced activation ferritic/martensitioc steels (RAFMs) has been finely tuned in order to increase the amount of fine MX precipitates and reduce coarse M$_{23}$C$_{6}$ carbides through the application of thermomechanical treatments. The microstructural investigations by TEM/STEM and EELS have shown M$_{23}$C$_{6}$, M$_{2}$X (Cr$_{2}$N), and MX (Ta, V, and Ti-rich) precipitates after tempering at 750 ��C/2 h. Higher N contens (0.04���0.03 wt%) seems to favored M$_{2}$X precipitation over MX with V contents round 0.2 wt%. MX-Ti rich presents sizes larger than MX (Ta or V rich). EELS anlysis have shown that the MX (Ta, V and Ti rich) precipitated after tempering at 750 ��C/2 h are mainly carbo-nitrides. Composition, size, number density and carbon and nitrogen content on MX and M$_{2}$X is discussed in terms of the composition of each Alloy.

A major part in the EUROfusion materials research program is dedicated to characterize and quantify nuclear fusion specific neutron damage in structural materials. While the majority of irradiation data gives a relatively clear view on the displacement damage, the effect of transmutation – i.e. especially hydrogen and helium production in steels – is not yet explored very well. However, few available results indicate that EUROFER-type steels will reach their operating limit as soon as the formation of helium bubbles reaches a critical amount or size. At that point, the material would fail due to embrittlement at the considered load. This paper presents a strategy for the mitigation of the before-mentioned problem using the following facts: • the neutron dose and related transmutation rate decreases quickly inside the first wall, that is, only a plasma-near area is extremely loaded • nanostructured oxide dispersion strengthened (ODS) steels may have an enormous trapping effect on helium and hydrogen, which would suppress the formation of large helium bubbles • compared to conventional steels, ODS steels show improved irradiation tensile ductility and creep strength In summary, producing the plasma facing, highly neutron and heat loaded part of blankets by an ODS steel, while using EUROFER97 for everything else, would allow a higher heat flux as well as a longer operating period. Consequently, we (1) developed and produced 14 % Cr ferritic ODS steel plates. (2) We fabricated a mockup with 5 cooling channels and a plated first wall of ODS steel, using the same production processes as for a real component. And finally, (3) we performed high heat flux tests in the HELOKA facility (Helium Loop Karlsruhe at KIT) applying short and up to 2 h long pulses, in which the operating temperature limit for EUROFER97 (i.e., 550 °C) was finally exceeded by 100 K. Thereafter, microstructure and defect analyses did not reveal defects or recognizable damage. Only a heat affected zone in the EUROFER/ODS steel interface could be detected. This demonstrates that the use of ODS steel could make a decisive difference in the future design and performance of breeding blankets.

Neutron multiplier materials are essential for self-sufficient tritium production and are closing the fuel cycle of fusion reactors. Until now, a concept of pebble bed consisting of interchanging layers of beryllium and lithium ceramic pebbles was considered for the Helium-Cooled Pebble Bed (HCPB) tritium-breeding module of the first experimental fusion reactor ITER as well as for the next demonstration fusion reactor DEMO. However, this concept depends on the availability of large amounts of pure beryllium pebbles and is also limited by its material properties like for example the tritium accumulation under irradiation. The results of tritium retention and analytical microstructural studies of beryllium pebbles obtained within the framework of the HIDOBE irradiation campaign suggest that a significant fraction of generated tritium (up to 100% below 500°C) is trapped within helium bubbles. Being negligible in the ITER tritium-breeding module (TBM), the total accumulated tritium inventory imposes severe safety issues and exceeds acceptable limits for the DEMO blanket. Therefore, advanced neutron multiplier materials such as beryllides have to be well characterized for their applicability in the HCPB blanket of DEMO and beyond. The usage of an advanced material with lower volumetric swelling, lower tritium retention, increased irradiation and chemical resistance as well as with higher melting temperature allows to switch from the pebble bed concept to a solid hexagonal block-based one. For the fabrication of titanium beryllide samples both the semi-industrial fabrication route utilizing the hot extrusion of rods and the industrial approach using vacuum hot pressing of a Be-Ti powder mixture were explored. In this contribution, we discuss reasons for the transition from pure beryllium to beryllides, respective changes of the HCPB blanket design and a successful demonstration of the feasibility of a beryllide block fabrication by an industrial method.

In future nuclear fusion reactors, plasma facing components have to sustain specific neutron damage. While the majority of irradiation data provides a relatively clear picture of the displacement damage, the effect of helium transmutation is not yet explored in detail. Nevertheless, available results from simulation experiments indicate that 9%-chromium steels will reach their operating limit as soon as the growing helium bubbles extent a critical size. At that point, the material would most probably fail due to grain boundary embrittlement. In this contribution, we present a strategy for the mitigation of the before-mentioned problem using the following facts. (1) The neutron dose and related transmutation rate decreases quickly inside the first wall of the breeding blankets, that is, only a plasma-near area is extremely loaded. (2) Nanostructured oxide dispersion strengthened (ODS) steels may have an enormous trapping effect on helium, which would suppress the formation of large helium bubbles for a much longer period. (3) Compared to conventional steels, ODS steels also provide improved irradiation tensile ductility and creep strength. Therefore, a design, based on the fabrication of the plasma facing and highly neutron and heat loaded parts of blankets by an ODS steel, while using EUROFER97 for everything else, would extend the operating time and enable a higher heat flux. Consequently, we (i) developed and produced 14%Cr ferritic ODS steel plates and (ii) optimized and demonstrated a scalable industrial production route. (iii) We fabricated a mock-up with five cooling channels and a plated first wall of ODS steel, using the same production processes as for a real component. (iv) Finally, we performed high heat flux tests in the Helium Loop Karlsruhe, applying a few hundred short and a few 2 h long pulses, in which the operating temperature limit for EUROFER97 (i.e. 550 ◦C) was finally exceeded by 100 K. (v) Thereafter, microstructure and defect analyses did not reveal critical defects or recognizable damage. Only a heat affected zone in the EUROFER/ODS steel interface could be detected. However, a solution to prohibit the formation of such heat affected zones is given. These research contributions demonstrate that the use of ODS steel is not only feasible and affordable but could make a decisive difference in the future design and performance of breeding blankets.

The development of appropriate materials for fusion reactors that can sustain high neutron fluencies at elevated temperatures remains a great challenge. Tungsten is one of the promising candidate materials for plasma-facing components of future fusion reactors, due to several favorable properties as for example a high melting point, a high sputtering resistivity, and a low coefficient of thermal expansion. The microstructural details of a tungsten sample with a 1.25 dpa (displacements per atom) damage dose after irradiation at 800°C were examined by transmission electron microscopy (TEM). Three types of radiation-induced defects were observed, analyzed and characterized: (i) voids with sizes ranging from 10 nm to 65 nm, (ii) dislocation loops with a size of up to 10 nm and (iii) W-Re-Os containing σ- and χ-type precipitates. The distribution of voids as well as the nature of the occurring dislocation loops were studied in detail. In addition, nano-chemical analyses revealed that the σ- and χ-type precipitates, which are sometimes attached to voids, are surrounded by a solid solution cloud enriched with Re. For the first time the crystallographic orientation relationship of the σ- and χ-phases to the W-matrix was specified. Furthermore, electron energy-loss spectroscopy could not unambiguously verify the presence of He within individual cavities.

In this study, we performed fracture toughness characterization of ten neutron-irradiated Eurofer97 variants using precracked miniature multi-notch bend bar (M4CVN) specimens based on the Master Curve method in the ASTM E1921 standard. The neutron irradiation was performed in the flux trap position of the High Flux Isotope Reactor (HFIR) of the Oak Ridge National Laboratory (ORNL) with the nominal irradiation temperature of 300°C and irradiation dose of 2.5 displacements per atom (dpa). Depending on the irradiation temperature and materials, we observed different degrees of irradiation hardening and embrittlement for ten Eurofer97 variants. The upper shift in the Master Curve reference temperature T0Q vs. the increase in Vickers microhardness values showed a liner relationship for only a few materials indicating different irradiation responses of the Eurofer97 variants.

A detailed microstructural analysis is one key factor for establishing structure–property relationships, which themselves are essential for manufacturing any device or part thereof. In particular, this paper focuses on the microstructural analysis of tungsten composite materials produced by powder injection molding (PIM). Our combined scanning electron microscopy (SEM) and transmission electron microscopy (TEM) approach revealed that W/TiC and W/Y2O3 composites are promising candidates for e.g. plasma facing components in future fusion reactors. The grains size distribution of all present phases was a log-normal one. TiC and Y2O3 precipitates in contrast to HfC ones limited the grain growth of the tungsten matrix during sintering about three times more efficient. The precipitate grain size was for all samples in the range of 1.7 µm–3.5 µm. Chemical interaction was only observed for TiC-based composites in the form of W diffusion into the TiC precipitate forming a mixed (Ti, W) carbide retaining the face-centered cubic (fcc) based crystal structure of pure TiC. The tungsten content in Y2O3 and HfC precipitates was found to be negligible. La2O3 was only observed in TEM attached to (Ti, W)C particles in the form of about 100 nm sized precipitates. As result, the Y2O3 and TiC containing samples are considered as promising materials for further detailed mechanical and microstructural investigations.

9%Cr reduced activation steels are the designated structural materials for future fusion reactors. The improvement of this class of alloys, especially for the extension of the operation limits, is the present scope of the EUROfusion materials project for advanced steels. Within this programme, new alloys are designed and fabricated to overcome some of the limitations of EUROFER97. In the present study, four 9%Cr alloys with some variations in the chemical compositions are compared to standard EUROFER97 batches. The main focus lies in the possible extension of the operation window to both higher and lower temperatures. The limits of the mechanical properties, which can be achieved through different heat and thermo-mechanical treatments, are explored within this work. The thermo-mechanical treatments used within this work consist of rolling within the austenite regime and rapid (water-) quenching to room temperature. Toughness from Charpy impact tests, creep to rupture lifetime and tensile strength are compared for the different treatments. EBSD and STEM investigations of the formed microstructures complete the presented work. Strong effects of the heat-treatments on the toughness and strength were observed. As expected, hardening the materials beyond the conventional treatment (980 °C/30 min + 750 °C/2 h) showed a major increase in tensile strength and decrease in toughness. Moderate variations in the alloy compositions open the possibility for extended temperature windows for these unconventional heat treatments (e.g. higher tempering temperature). Thermo-mechanical treatments are effective to improve hardening and other properties by modifications of the distribution of secondary phases. As expected, the different heat treatments also showed the well-known effects on the mechanical properties (e.g. hardness and strength). When compared, the effects of heat treatments on the mechanical properties are far beyond the influence of minor changes in the alloy compositions.

Research and development projects on materials for applications exposed to high temperatures and radiation include comprehensive mechanical and metallographic examinations as well as analysis methods based on electron microscopy investigations using SEM (scanning electron microscope) and TEM (transmission electron microscope). In addition to 2D EDS and EBSD analyses in the μm range, the imaging 3D analysis method has become more and more important in the past few years. Particularly, the combination of SEM images with simultaneously obtained EDS and EBSD data during 3D analyses is a test method which complements previous material testing and produces clear results since much larger data sets are used in order to determine the composition of a material and the phases present in its microstructure. Since 2014, our institute has been working with a workstation (FIB/SEM) from the company Zeiss, a crossbeam AURIGA 40 with 3D analysis. This paper will present unparalleled results of the investigations carried out on the structural and high-temperature material 12Cr0.36Ta steel, nanocrystalline tungsten and intermetallic NiAl-Cr alloys using the above-mentioned 3D technique.

This work has been carried out within the framework of the German Helmholtz Association and has received funding from the topic “Materials Research for the Future Energy Supply”. The work of M. Parish and Rainer Ziegler is gratefully acknowledged. Thanks are also due to the team of the chemical laboratory at the KIT for performing the chemical analysis. The help of the beamline staff at ELETTRA (project 20140052 ) synchrotron radiation facility is acknowledged. We acknowledge support by Deutsche Forschungsgemeinschaft and Open Access Publishing Fund of Karlsruhe Institute of Technology., Austenitic oxide dispersion strengthened (ODS) steels are one of the candidates as a structural material for high-temperature applications in future power plants. To guarantee the necessary high production yield, the production process was improved in terms of reproducibility and scalability, by adding a process control agent (PCA) during the milling process. Due to this addition and the inherent change of the production process, the produced powder was thoroughly investigated using transmission electron microscopy and X-ray absorption spectroscopy methods to reveal the formation of chromium-rich carbides adjunct to titanium. Hence, less titanium was available to form the preferred complex nano-oxides the addition of carbon to the system influences the formation of precipitates severely in terms of their amount and size. The mechanical alloying process itself was unaffected by the addition of a PCA, and mixing and alloying of used elements still occurs., Deutsche Forschungsgemeinschaft; Karlsruhe Institute of Technology; Helmholtz Association; Institute of Solid State Physics, University of Latvia as the Center of Excellence has received funding from the European Union’s Horizon 2020 Framework Programme H2020-WIDESPREAD-01-2016-2017-TeamingPhase2 under grant agreement No. 739508, project CAMART²

Ten Eurofer97 steel variants, produced by non-standard fabrication-processing routes and modified alloying chemistries, were studied by neutron irradiations in the high flux isotope reactor. The irradiations were performed to ITER-TBM relevant conditions of ∼255–350 °C, 2.94–3.24 dpa. We quantified the irradiation-induced degradation of the steels using mechanical property tests. All the steels suffered from irradiation hardening, where a significant increase in Vickers microhardness and yield stress (σYS) occurred, accompanied with severe loss of tensile elongation. The extent of hardening was material dependent. For Tirr = 300±30 °C, most steels showed σYS increase in the range of ∼30% to as high as ∼66%, except for a low temperature tempered steel with σYS increase below 15%. Despite large losses in elongation, most failures were ductile. Significant post-necking ductility was retained with reduction in area (RA) between 65–75%, but

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 W f /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.

Conventionally produced tungsten (W) sheets are brittle at room temperature. In contrast to that, severe deformation by cold rolling transforms W into a material exhibiting room-temperature ductility with a brittle-to-ductile transition (BDT) temperature far below room temperature. For such ultrafine-grained (UFG) and dislocation-rich materials, the mechanism controlling the BDT is still the subject of ongoing debates. In order to identify the mechanism controlling the BDT in room-temperature ductile W sheets with UFG microstructure, we conducted campaigns of fracture toughness tests accompanied by a thermodynamic analysis deducing Arrhenius BDT activation energies. Here, we show that plastic deformation induced by rolling reduces the BDT temperature and also the BDT activation energy. A comparison of BDT activation energies with the trend of Gibbs energy of kink-pair formation revealed a strong correlation between both quantities. This demonstrates that out of the three basic processes, nucleation, glide, and annihilation, crack tip plasticity in UFG W is still controlled by the glide of dislocations. The glide is dictated by the mobility of the screw segments and therefore by the underlying process of kink-pair formation. Reflecting this result, a change of the rate-limiting mechanism for plasticity of UFG W seems unlikely, even at deformation temperatures well below room temperature. As a result, kink-pair formation controls the BDT in W over a wide range of microstructural length scales, from single crystals and coarse-grained specimens down to UFG microstructures.

Different manufacturing routes are investigated at the KIT INR for the realization of First Walls (FW) and internal components for the Helium Cooled Pebble Bed (HCPB) nuclear fusion reactor component Breeding Blanket (BB). As result of cost estimation studies in 2017 and 2018, Additive Manufacturing (AM) has been identified to provide promising options beyond conventional technologies. Selective Laser Melting (SLM) and Direct Energy Deposition (DED) applied as Cold Spray (CS) were addressed. SLM was followed with focus on the development of BB internal structures. CS turned out as a very goal oriented and effective fabrication process with regard to the First Wall geometrical requirements. Further, a large potential for cost reduction compared to other technologies was identified. Based on a variation of the DED process CS applied in alternation with machining and an option for temporary filling cavities, an extended process chain was developed in collaboration with industry. This process chain also enables Laser structuring of internal channel surfaces for heat transfer enhancement structures. This paper summarizes the developments in terms of application of AM for the HCPB BB concept where SLM is focused as an option for BB internal structures assembled as hybrid components and CS addressed for fabrication of the First Wall.