Wall Of Future Fusion Reactors Research Articles

Processing and Properties of Sintered W/Steel Composites for the First Wall of Future Fusion Reactor

Functionally graded tungsten/steel composites are attractive to be used as an interlayer to join tungsten (W) and steel for the first wall of future fusion reactor to reduce the thermally induced stresses arising from the different coefficient of thermal expansion (CTE) of W and steel. W/steel composites, with three W contents: 25, 50 and 75 vol% W, will serve as individual sublayers of this functionally graded material. Therefore, the present work exploits an emerging sintering technique, field-assisted sintering technology, to produce these composites. Firstly, a systematic parameter study was conducted aiming to reduce the residual porosity to a minimum while keeping the formation of intermetallic phases at the W/steel interface at a low level. The optimized composites 25, 50 and 75 vol% W achieved a relative density of 99%, 99% and 96%, respectively. Secondly, mechanical tests at elevated temperatures reveal that these composites are ductile above 300 °C, which is the minimum operating temperature of the first wall. Lastly, the measured CTE, specific heat capacity and thermal conductivity were consistent with the theoretically expected values.

The effect and mechanism of closed double equal channel angular pressing deformation on He+ irradiation damage of low activation steel

As the preferred structural material for the cladding first wall of future fusion reactor, improving radiation tolerance of low activation steel is a key problem to ensure nuclear safety. Here we used Closed Double Equal Channel Angular Pressing to produce ultra-fine grain low activation steel and performed He+ irradiation at 500°C on initial and ultrafine-grained low activation steel. It was found that the martensite lath and precipitated phases were effectively refined after deformation. The ultrafine-grained low activation steel reduced the size and density of He bubbles, defect-clusters and dislocation loops generated by irradiation. Deformation created high angle boundaries and facilitated the precipitation of the precipitation phases, leading to increase defect sinks, which consequently enhance the radiation tolerance of low activation steel. Furthermore, since irradiation-induced defects tend to be distributed along the boundaries, the ultra-fine grain structure facilitates the homogenization of irradiation damage. This provides a new reference for the design of irradiation-resistant low activation steel.

Atmospheric plasma spraying of functionally graded steel/tungsten layers for the first wall of future fusion reactors

Functionally graded steel/tungsten layers may be used as interlayers in the first wall of future fusion reactors to balance thermally-induced stress peaks in the steel‑tungsten joint. In this work, a modified water-stabilized atmospheric plasma spraying set-up is used to deposit uniform and functionally graded steel/tungsten coatings at elevated substrate temperatures. Uniform coatings were used to characterise individual sublayers of graded coatings in detail. The thermal expansion, thermal conductivity, Young's modulus and yield strength of the layers are promising for the application in steel‑tungsten joints and can be explained by microstructural observations. Only at a substrate preheating temperature of 900 °C the formation of intermetallic precipitates during deposition was observed.

Microstructure of the tungsten and reduced activation ferritic-martensitic steel joint brazed with an Fe-based amorphous alloy

The effective joining of tungsten (W) and reduced activation ferritic-martensitic (RAFM) steels is crucial to fabrication of the divertors and first wall of future fusion reactors. In the present work, a low-activation Fe-based amorphous alloy of Fe67.8Cr11.5Si2.1B18.6 (at. %) is designed as the filler metal for brazing W and RAFM steels. Crack-free joint has been achieved with the amorphous alloy filler and a vanadium interlayer metal by short-time vacuum brazing at 1270 °C. Layer structures composed of intermetallic phases and single solid-solution phases, respectively, are alternately formed in the W/RAFM steel joint, exhibiting alternating hard and soft mechanical characteristic. Long-distance diffusion of W atoms to the steel substrate is blocked by the formation of FeW2B2 and Fe3B phases, and Si and B elements are confined within the main bonding seam of the joints. Microstructure recovery for the steel is realized following the standard heat treatment procedures. The present results suggest a promising way of making strong and tough W/RAFM joints with low-activation Fe-based amorphous alloys.

Ultra-fast sintered functionally graded Fe/W composites for the first wall of future fusion reactors

Aiming at the realisation of nuclear fusion reactors, the joining of W and steel parts is currently examined. Based on proposals to implement functionally graded steel/W materials (FGMs) in the joint to cope with thermal stresses, the present contribution introduces a novel fabrication method for steel/W FGMs. Electro Discharge Sintering (EDS) was used to consolidate Fe/W powders within milliseconds at atmosphere. Due to the short process time, the formation of detrimental intermetallic Fe-W precipitates is limited compared to established fabrication methods.The current work first presents results of the Fe/W powder processing, then a feasibility study regarding the fabrication of homogeneous and graded Fe/W composites with W volume fractions of 0, 25, 50 and 75 % via EDS is presented. Lastly, the composites are characterised microstructurally, thermo-physically, and mechanically in detail.

On Oxidation Resistance Mechanisms at 1273 K of Tungsten-Based Alloys Containing Chromium and Yttria

Tungsten (W) is currently deemed the main candidate for the plasma-facing armor material of the first wall of future fusion reactors, such as DEMO. Advantages of W include a high melting point, high thermal conductivity, low tritium retention, and low erosion yield. However, was an accident to occur, air ingress into the vacuum vessel could occur and the temperature of the first wall could reach 1200K to 1450K due to nuclear decay heat. In the absence of cooling, the temperature remains in that range for several weeks. At these temperatures, the radioactive tungsten oxidizes and then volatilizes. Smart W alloys are therefore being developed. Smart alloys are supposed to preserve properties of W during plasma operation while suppressing tungsten oxide formation in case of an accident. This study focuses on investigations of thin film smart alloys produced by magnetron sputtering. These alloys provide an idealistic system with a homogeneous distribution of the elements W, chromium (Cr), and yttrium (Y) on an atomic scale. The recommended composition is W with 12 weight % of Cr and 0.5 weight % of Y. Passivation and a suppression of WO3 sublimation is shown. For the first time, the mechanisms yielding the improved oxidation resistance are analyzed in detail. A protective Cr2O3 layer forms at the surface. The different stages of the oxidation processes up to the failure of the protective function are analyzed for the first time. Using 18O as a tracer, it is shown for the first time that the oxide growth occurs at the surface of the protective oxide. The Cr is continuously replenished from the bulk of the sample, including the Cr-rich phase which forms during exposure at 1273K. A homogenous distribution of yttria within the W-matrix, which is preserved during oxidation, is a peculiarity of the analyzed alloy. Further, an Y-enriched nucleation site is found at the interface between metal and oxide. This nucleation sites are deemed to be crucial for the improved oxidation resistance.

Oxidation resistance of bulk plasma-facing tungsten alloys

Tungsten (W) currently is the main candidate as plasma-facing armour material for the first wall of future fusion reactors, like DEMO. Advantages of W include a high melting point, high thermal conductivity, low tritium retention, and low erosion yield. However, in case of an accident, air ingress into the vacuum vessel can occur and the temperature of the first wall can reach 1200 K to 1450 K due to nuclear decay heat. In the absence of cooling, the temperature will remain in that range for several weeks. At these temperatures the radioactive tungsten oxide volatilizes. Therefore, ‘smart’ W alloys are developed that aim to preserve the properties of W during plasma operation coupled with suppressed tungsten oxide formation in case of an accident.This study focusses on oxidation studies at 1273 K of samples produced by mechanical alloying followed by field assisted sintering. In a first step the sintering is optimized for tungsten (W) – chromium (Cr) -yttrium (Y) alloys. It is shown that the best oxidation resistance is achieved with submicron grain sizes. This yields a closed, protective oxide layer. In a second step the influence of the grinding process during sample preparation is analysed. It is shown that scratches initiate failure of the protective oxide. In a third step the oxidation and sublimation is measured for weeks – for the first time the sublimation is directly measured in order to determine the potential hazard in comparison to pure W. It is shown that the oxidation is suppressed in comparison to pure W. However, sublimation at a rate of 1×10−6mgcm−2s−1 starts after a few days. Nevertheless, the progess in smart alloys is evident: sublimation is delayed by about two days and complete mechanical destruction of the first wall is avoided.

Atomistic simulations of deuterium irradiation on iron-based alloys in future fusion reactors

Iron-based alloys are now being considered as plasma-facing materials for the first wall of future fusion reactors. Therefore, the iron (Fe) and carbon (C) erosion will play a key role in predicting the life-time and viability of reactors with steel walls. In this work, the surface erosion and morphology changes due to deuterium (D) irradiation in pure Fe, Fe with 1% C impurity and the cementite, are studied using molecular dynamics (MD) simulations, varying surface temperature and impact energy. The sputtering yields for both Fe and C were found to increase with incoming energy. In iron carbide, C sputtering was preferential to Fe and the deuterium was mainly trapped as D2 in bubbles, while mostly atomic D was present in Fe and Fe–1%C. The sputtering yields obtained from MD were compared to SDTrimSP yields. At lower impact energies, the sputtering mechanism was of both physical and chemical origin, while at higher energies (>100eV) the physical sputtering dominated.

Hydrogen permeation through disordered nanostructured tungsten films

We report results on long-term hydrogen outgassing and permeation studies of structurally highly disordered tungsten films, deposited on 40mm diameter highly permeable Eurofer substrates, using the Pulsed Laser Deposition technique. Hydrogen interaction with tungsten is becoming a highly relevant topic since tungsten was recognized as the most promising candidate for the first wall of future fusion reactors. Prediction of hydrogen isotopes migration and their abundance after plasma exposure is uncertain due to a great role played by structural disorder that is formed on the W surface. Vacancy sites are theoretically predicted to trap multiple H atoms exothermically, but their density and their potential influence on permeability has not been experimentally investigated yet. In our work, permeability of W films having different thicknesses (1 and 10μm) was initially extremely low, and was gradually increasing over a several-day campaign. The final values at 400°C, lying between P=1.46×10−15molH2/(msPa0.5) and P=4.8×10−15molH2/(msPa0.5), were substantially lower than those known for well ordered films. Surprisingly, the 10μm thick W film initially contained a very high amount of hydrogen, ∼0.1H/W, which was gradually releasing during the twenty-day campaign.

Interface reactions and control of diffusion at the interface between SiC fibres and layer of deposited Fe–9Cr base alloy

The ferromagnetic 9Cr steel EUROFER 97 is considered as a structural material for the first wall of future fusion reactors. To solve a problem of limitation of the maximum operation temperature of 550 °C for EUROFER 97, a new metal matrix composite SiC/EUROFER 97 is proposed. To create a thermally stable interface between the SiC fibres and the steel matrix, the following interlayers were tested as a diffusion barrier: Cr, Re, W and Re/W. Samples with these interlayers were investigated before and after annealing at 750 °C which is supposed to be the operation temperature of the composite material. To simplify examination of the coatings using available analysis methods (SEM, IBA and XRD), the fibre/steel system was approximated by planar samples with identical layer structure. The experiments with both SiC fibres and planar substrates revealed W to be an efficient diffusion barrier.

Synchrotron, neutron and classical X-ray measurement in an electron beam welded martensitic steel

Abstract The F82-H modified martensitic steel is currently one of the best options for the first wall of future fusion reactors. Martensitic steel samples of F82-H (0.1 C; 8 Cr; 2 W, 0.2 V; 0.1 Mn; 0.1 Si wt %) have been electron beam welded. Strain analysis by classical X-ray, synchrotron and neutron diffraction were carried out at different distances from the weld centre (0, 1, 2, 3, 4 and 5 mm). Strain and stress calculations require the determination of the exact unstressed lattice parameter (d 0), which is determined by XRD. Material characterization and line broadening measurements by X-rays and neutrons were performed. The integral breadth values measured by X-rays and neutrons coincide and can be compared to the micro-hardness profile.

Ferritic/martensitic dual-phase steel as fusion reactor materials

In ferritic/martensitic steels, the swelling rate is much lower than in austenitic steels. Therefore, ferritic/martensitic steels may become the next candidate alloys for the first wall of future fusion reactors. Until now, the martensitic steels HT-9 and modified 9Cr-1Mo steel have been investigated. These steels have excellent high-temperature creep properties but are not necessarily better in Ductile Brittle Transition Temperature (DBTT) properties than austenitic steel. Especially their weldability is a problem. For these reasons new ferritic/martensitic dual-phase steels have been considered. In this presentation, the advantages and disadvantages in their use as fusion reactor materials will be discussed. The recommended compositions are mainly low-C 9Cr-1 or 2Mo including some modifications. Since these alloys have wider two-phase regions than 9Cr-1Mo steel, after normalizing, a ferritic/martensitic structure with a fixed percentage of ferrite can be obtained. The mechanical properties, weldability and other properties are compared between single- and dual-phase steels. And the characteristics of the ferritic/martensitic dual phase steels are introduced.