Wire Arc Additive Manufacturing (WAAM) of 316L stainless steel typically results in columnar grains, high dislocation densities, and residual porosity, which limit toughness compared to conventional material. This study evaluates Hot Isostatic Pressing (HIP) as a post-processing route to refine the microstructure and eliminate defects by investigating four HIP cycles (1000-1200 degrees C, 100-150 MPa) using EBSD-EDS, neutron diffraction, tomography, and mechanical testing. Increasing HIP temperature and pressure promoted dislocation recovery and recrystallisation while dissolving metastable delta-ferrite; however, processing at 1000 degrees C induced brittle sigma-phase formation, while the 1200 degrees C/150 MPa cycle (HIP-4) produced a fully recrystallised, chemically homogeneous austenitic structure. HIP-4 reduced porosity by 98.8 % and restored a mechanical response comparable to conventionally processed 316L, characterised by improved ductility and strain-hardening capacity despite a reduction in yield strength. Ultimately, HIP-4 establishes an optimal post-processing window for achieving concurrent densification and microstructural homogenisation, significantly enhancing the mechanical performance and reliability of WAAM 316L components.

Wire-Arc Additive Manufacturing (WAAM) enables the cost-effective production of large-scale steel structures but often introduces significant residual stresses that can compromise structural integrity. This study investigates the residual stress distribution in WAAM-produced 316 L stainless steel components using neutron diffraction and contour method, with particular focus on cross-correlating those two independent techniques. Residual stress measurements from both methods show strong agreement, providing valuable input for finite element models (FEM) of the process. Neutron diffraction enables non-destructive subsurface and bulk measurements, but its accuracy heavily depends on the preparation of stress-free reference samples. A detailed analysis of these reference samples revealed a strong crystallographic texture, though the overall variation remained small across the height of the WAAM component. The contour method, in contrast, is unaffected by texture or chemical composition variations, offering a key advantage over diffraction-based techniques. However, as a destructive method, it is limited to measuring residual stress in a single direction. This study discusses the strengths and limitations of both techniques in the context of WAAM, particularly regarding spatial resolution, reference sample requirements, and applicability to large-scale components. By refining residual stress measurement methodologies and identifying key constraints, this study advances the approach to residual stress assessment of WAAM printed structures and hence contributes to improved process control and enhanced structural reliability of additive manufacturing for civil and structural applications.

An, as cast, VCrFeW0.2 refractory medium entropy alloy (RMEA) was designed for fusion reactor divertor applications, focusing on reduced cost, low activation and compositional stability (low transmutation rates). The as-cast alloy was irradiated to a fluence of 5.6 x 10(17) ions/cm(2) at room temperature with 5 MeV helium ions whose energy have been uniformly attenuated to 0.4 MeV and 5 MeV via energy degradation device prior to sample irradiation. Pre and post irradiation, its mechanical properties were evaluated micro-tensile testing. Prior to irradiation, the VCrFeW0.2 alloy demonstrated good strength and ductility, with a yield strength of 1464 MPa and strain to UTS (epsilon(UTS)) of 4.6 %, maintaining comparable strength to pure tungsten (1403 MPa) but with greater strain to UTS (1.3 %). Post irradiation, the VCrFeW0.2 alloy exhibited remarkable damage resistance; its strength increased by only similar to 160 MPa, and it retained strain to UTS with a (epsilon(UTS)) of 2.9 %. It performed better than pure tungsten tested under identical irradiation conditions where there was similar to 1800 MPa increase in yield strength and a complete loss of plasticity. The micro-tensile results were supported by nanoindentation tests and Vickers hardness testing was also undertaken to show the yield strength values are representative of macro scale, bulk behavior. TEM and comparison with existing literature on RMEA/RHEA are presented here to understand the reason for difference in performance between VCrFeW0.2 alloy and pure tungsten.

The microstructure and high-temperature corrosion behavior of Hastelloy C276 fabricated by wrought processing and wire arc additive manufacturing (WAAM) were systematically investigated. The wrought alloy exhibited equiaxed grains with uniform elemental distribution, and sparse nanoscale grain boundary precipitates. In contrast, WAAM specimens taken from the top, middle, and bottom regions displayed a textured dendritic gamma-Ni matrix with interdendritic Mo/Cr segregation and scattered second-phase particles. Rapid cooling near the substrate produced cellular structures in the bottom region, while limited heat transfer from subsequent layers led to finer and shorter dendrites in the top region. Corrosion testing in FLiNaK at 750 degrees C for 500 h resulted in average corrosion rates of 0.0935 mm/year (wrought) and 0.0979 mm/year (WAAM). Post-corrosion characterization and modeling identified chromium depletion as the primary degradation mechanism, driving vacancy formation, dynamic recrystallization, and localized pitting, particularly in defect-rich interdendritic and grain-boundary regions. A Cr-depletion-based damage factor was integrated into a corrosion model to predict mass loss and corrosion rate. Although Mo segregation and dynamically formed topologically close packed phases locally hindered Cr diffusion, their overall influence on corrosion depth was limited. The WAAM alloy exhibited slightly inferior corrosion resistance relative to wrought C276, attributable to its higher interdendritic area fraction, while the extent of corrosion remained governed by the Cr-depletion-controlled dissolution process.

Uranium oxide hydrate frameworks (UOHFs) with lanthanide ions are fundamentally important to uranium crystal chemistry. Earlier works revealed a possible phase transition from high symmetry space groups (monoclinic or orthorhombic) for UOHF-Ln (Ln = Pr-* Dy) to a low symmetry space group (triclinic) for UOHF-Ln (Ln = Er-* Lu). However, given holmium lies close to this apparent transition boundary, whether these frameworks with holmium adopt a high or low symmetry structure is unknown. Herein, we report the synthesis of Ho2(H2O)4(OH)2[(UO2)10UO13(OH)4]& sdot;H2O (UOHF-Ho) and subsequent structural and spectroscopic investigations. Synchrotron single crystal X-ray diffraction confirmed that UOHF-Ho crystallises in the orthorhombic C2221 space group, further validated by electron diffraction with transmission electron microscopy. The framework structure is constructed with beta-U3O8 type layers linked by double uranyl units in pentagonal bipyramids, with disordered Ho3+ ions lying inside the channels. While Raman spectroscopy revealed the U6+ dominant vibrational modes, diffuse reflectance spectroscopy unearthed characteristic absorption bands for both U6+ and Ho3+ ions. This work uncovers the exact phase transition for UOHF-Ln series and has implications to the uranium structural chemistry and possible spent nuclear fuel alterations.

Uranium hydrolysis processes and associated products are fundamentally important especially to uranium geochemistry and the nuclear fuel cycle. Herein, we report the formation and characterisation of a new electroneutral uranium oxide hydrate framework (UOHF), U(H2O)8[(UO2)10UO13(OH)4] (UGHF2), using both structural and spectroscopic techniques. As a minor phase from the hydrothermal uranium hydrolysis in the presence of Cu(II) ions, compound UGHF2 crystallises in the triclinic P1 space group having a framework structure constructed with beta-U3O8 type layers pillared by double pentagonal uranium bipyramids and tetravalent U(IV) ions sitting inside the framework channels. The crystal structure has been revealed using synchrotron single crystal X-ray diffraction and confirmed by electron diffraction with transmission electron microscopy. The characteristic vibrational modes due to the presence of various hexavalent uranium centres were revealed by Raman spectroscopy. The successful synthesis and characterisation of UGHF2, together with an earlier reported U(H2O)2[(UO2)10O10(OH)2(UO4)(H2O)2] (UGHF1), highlights the complex nature of uranium hydrolysis processes and its related products, with direct implications to fundamental uranium chemistry, geochemistry and the current nuclear fuel cycle.

Light, strong, and radiation-tolerant materials are essential for advanced nuclear systems and aerospace applications. However, the comprehensive properties of current radiation-tolerant materials are far from being satisfactory in harsh operating environments. In this study, a high-throughput–designed NbVTaSi refractory eutectic medium entropy alloy realizes the controllable formation of the β-Nb 5 Si 3 phase with a high content and has outstanding comprehensive properties, i.e., lightweight, high yield strengths at room temperature and 850°C, and excellent He-ion irradiation resistance. According to density functional theory calculations and experimental findings, the prefabricated lattice distortion of the Nb 50 V 42 Ta 8 phase leads to great phase stability under severe He-ion irradiation conditions, while the dual characteristics of the semi-coherent interface and hyperstatic lattice structure of the high-content β-Nb 5 Si 3 phase dominate its outstanding He-ion irradiation resistance. This study sheds light on the design strategy for comprehensive properties and development of future radiation-tolerant materials for advanced nuclear systems and aerospace applications.

An international collaboration was established as a Coordinated Research Project (CRP) under the IAEA and entitled Accelerator Simulation and Theoretical Modelling of Radiation Effects-II (SMoRE-II). It was created to determine, by way of a Round Robin process, the degree to which ion irradiations produced the same irradiated microstructure when conducted in different labs on the same alloy and provided with the same irradiation protocol. The Round Robin consisted of 13 participating organizations from 9 IAEA member states with ion irradiations conducted at all CRP partner sites on samples of a single alloy (T91) from a single billet with the same thermal-mechanical history, and with a specific protocol for conducting the irradiations. Of the 14 parameters specified for the ion irradiations, only 1 of 12 facilities was able to follow the protocol exactly. Major differences included vacuum pressure, temperature measurement and control, beam mode (raster-scanning vs. steady beam), and dosimetry. The microstructure features characterized were the sizes and number densities of cavities, dislocation loops, precipitates, and the radiation induced segregation. While loop size and number density appeared to correlate with carbon content, no such correlation was identified for cavities. The divergence from the irradiation protocol undoubtedly affected the irradiated microstructure with carbon contamination occurring in most cases. The cavity, dislocation loop and precipitate microstructures all fell within the range of that in the literature. Additionally, a T91 sample that was irradiated to 47 dpa at 376 degrees C in the BOR-60 reactor was selected for comparison of the microstructure to those in the Round Robin study.

Understanding radiation damage resistance in Grade 91 steel (P91) is essential for the development of materials for future nuclear components. This study explores the combined effects of creep aging and helium ion irradiation on the microstructure and mechanical properties of P91 steel. Creep aging was conducted under a stress of 110 MPa at 625 degrees C for 475 h, followed by irradiation with 5 MeV helium ions to a fluence of 5.6 x 1017 ions/cm2, creating a uniform radiation-affected zone with an average damage level of 0.6 dpa. Microstructural changes and mechanical responses were assessed through detailed microstructural observations and nanoindentation, supported by finite element modelling. The results show that creep aging led to a slight reduction in hardness from 2.66 GPa to 2.45 GPa, primarily due to carbide coarsening. Significant irradiation hardening was observed, with hardness increasing by 87 % in the as-received condition and by 99 % in the creep-aged condition. A threedimensional finite element model was developed to reverse-engineer stress-strain relationship from nanoindentation load-displacement data. This study underscores the significant impact of combined creep aging and irradiation on P91 steel, with important implications for its use in nuclear applications.

Uranium oxide hydrate (UOH) materials with alkali metals are fundamentally important. We report new structural insights into UOH materials with K+/Rb+ ions, with five compounds being synthesised and characterised, revealing both layered structures and uranium oxide hydrate frameworks (UOHFs). Compound K (H2O)2.5[(UO2)3O2(OH)3] (U-K1) crystallises in the orthorhombic Pnnm space group with a layered structure composed of alpha-U3O8 layers and interlayer K+ ions. Compound K3(H2O)2[(UO2)10(UO4)O8(OH)5] (U-K2) crystallises in the monoclinic C2/c space group with a framework structure composed of beta-U3O8 layers pillared by double uranyl units in pentagonal bipyramids and K+ ions inside the framework channels. Although Rb(H2O) [(UO2)4O2(OH)5] (U-Rb1) crystallised in the trigonal R3 space group and Rb(H2O)2.5[(UO2)3O2(OH)3] (U-Rb2) crystallised in the orthorhombic Pnnm space group, both have layered structures with alpha-U3O8 type layers and interlayer Rb+ ions, differing in their U:Rb ratios: 4:1 for U-Rb1 and 3:1 for U-Rb2. Similar to U-K2, compound Rb3(H2O)2[(UO2)10(UO4)O8(OH)5] (U-Rb3) crystallises in the monoclinic C2/c space group having a framework structure with Rb+ ions inside the framework channels. This work accounts for the first report of UOHFs stabilised with exclusively alkali metal ions, highlighting the complexity of the UOH systems, with implications to uranium geochemistry and spent fuel alterations under geological disposal.

The microstructure and high-temperature creep mechanisms of Ni-based Hastelloy C276 superalloy fabricated using wire and arc-based directed energy deposition were investigated systematically and innovatively. The microstructural investigation revealed that the as-fabricated samples comprise γ-Ni matrix and topologically close-packed (TCP) P phase precipitates. The γ matrix subgrains and grains are spread over multiple highly textured columnar dendrites, with a majority of γ <001> crystallographic orientations closely aligned along the deposition direction. Moreover, the interdendritic regions exhibit severe Mo segregation, P phase particles, and dislocation bands. Creep tests were conducted on miniature samples under various temperature and stress conditions, loaded either in the deposition direction (DD) or travel direction (TD). DD samples exhibit lower minimum strain rates, greater strains-to-failure, and longer creep rupture lifetimes than TD samples, indicating significant creep anisotropy. Dislocation creep was identified as the primary creep mechanism for both DD and TD conditions. During creep, dynamic precipitation of TCP phases occurred in the interdendritic regions, resulting in varying creep resistance between interdendritic and dendritic core regions. Isostress and isostrain models, considering both crystallographic texture and precipitation strengthening, reasonably predicted the observed creep anisotropy during the secondary creep stage. Additionally, variations in the Schmid factor led to significant deformation incompatibility among dendrites in TD samples. Dislocation accumulation in TD sample interdendritic regions promoted new grain nucleation, triggering dynamic recrystallisation, facilitating grain boundary sliding, and accelerating tertiary creep. Furthermore, TCP phase particles in the interdendritic regions contributed to microcrack development, further accelerating creep fracture, especially in the TD condition.

Metal oxides possessing a large surface area, pore volume and desirable pore size provide more varieties and active industrial potentials. Nevertheless, it is very challenging to produce crystal metal oxides while keeping satisfactory porosity features, especially for ternary compositions. High temperature is usually needed to produce crystal metal oxides, which readily leads to the collapse of the pore structure. Herein, by employing a ‘soft’ dispersant agent and a hard silica template, ZrO2, TiO2 and Zr-Ti solid solutions having a tetragonal crystal structure are produced and the silica-leached materials are characterized from macroscopic to atomistic scales. The micron-sized particulate powders are composed of nanoscale ‘building blocks’, with crystallite sizes between ~8 and 21 nm. These polycrystalline ceramic powders exhibit a high specific surface area (up to ~200 m2·g−1) and pore volume (up to 0.5 cm3·g−1), with a pore size range of ~5–20 nm. Importantly, the Zr/Ti–O–Si–OH chemical bonds exist on the particle surface, with about two-thirds of the surface covered by silica. The hydroxyl groups can further post-graft organic ligands or directly associate with species. Synthesized mesoporous metal oxides are highly homogenous and could potentially be used in various applications because of their tetragonal structure and porosity features.

Refractory high‐entropy alloys (RHEAs) show potential for use in extreme environments, such as advanced nuclear reactors, owing to their high melting temperature, and often outstanding combinations of mechanical properties, corrosion resistance, and irradiation‐damage tolerance. This study evaluates the fracture toughness of a TiZrNbHfTa RHEA across different scales and microstructures, with a focus on the impact of He2+‐ion irradiation. Micro‐ and millimeter‐scale specimens with nanocrystalline (NC) microstructures are compared to existing ASTM standard sized coarse‐grained (CG) specimen data, with critical dimensions spanning over three orders of magnitude, from 10 μm to 12 mm. The ASTM standard sized CG specimens exhibit a fracture toughness 41‐fold greater than their NC microscale counterparts (210–5.1 MPa m1/2), while NC millimeter‐scale specimens show a 7.5‐fold higher fracture toughness than NC microscale specimens (38.1–5.1 MPa m1/2). He2+‐ion irradiation leads to a 27% decrease in fracture toughness in the NC microscale specimens. The results highlight the impact of sample dimensional scale, microstructure, and ion irradiation on the fracture toughness of the RHEA, indicating a need for thorough examination of such factors when investigating the mechanical properties of these materials.

Refractory high‐entropy alloys (RHEAs) are candidate structural materials for nuclear applications due to their promising high‐temperature mechanical performance and irradiation tolerance. However, most body‐centered cubic (BCC) RHEAs form additional phases depending on their thermal history, with few studies assessing their effect on irradiation tolerance. This study characterizes the impact of phase transformations on the room‐temperature irradiation tolerance of a nanocrystalline TiZrNbHfTa RHEA by assessing its microstructure and micromechanical properties before and after thermal treatments between 500 and 800 °C. The alloy demonstrates exceptional irradiation tolerance before and after 500 °C treatments for 1–100 h, which induce BCC to hexagonal close‐packed (HCP) phase transformation, with excellent microstructural stability and minimal irradiation‐induced hardening. Conversely, 800 °C treatment for 1 h forms two major BCC phases and a minor HCP phase, negatively impacting both pre‐ and post‐irradiation mechanical performance and causing significant irradiation‐induced hardening and embrittlement. Additionally, this research identifies a second HCP phase in the 500 °C, 100 h‐treated condition, marking its first mention in the literature. This study emphasizes the importance of assessing temperature and phase formation effects on the irradiation tolerance of RHEAs for future nuclear reactors.

Additive manufacturing (AM) enables the rapid fabrication of complex shapes using engineering materials such as austenitic stainless steels and can imbue them with high irradiation resistance for use in reactor components. This is attributed to their refined grain structure and high grain boundary area. In this study, austenitic stainless steels (type 304) fabricated via direct energy deposition (DED) and powder bed fusion (PBF) techniques were irradiated with 5 MeV He ions to an approximate dose of 0.6 dpa at 300°C; subsequently, they were characterized through electron microscopy and the micro tensile testing. The results revealed that austenitic stainless steels manufactured using AM methods exhibited outstanding mechanical performance. The high performance of austenitic stainless steels fabricated through the DED technique can be attributed to their high tensile strength and excellent ductility elongation. This excellent performance is believed to be caused by the low stacking fault energy and the corresponding martensite formation during deformation. In particular, it was found that better mechanical properties were maintained even after helium irradiation, which is an important result obtained from the micro-tensile test. Even a small variation in the chemical composition and sub-microstructure of AM materials could result in improved irradiation tolerances.

Carbon fibre-reinforced polyetheretherketone (CF-PEEK) composites have gained significant usage across diverse industries like automotive and aerospace due to their desirable characteristics. These properties encompass recyclability, low density, high strength, wear resistance and thermal stability. The components made from CF-PEEK composites for space applications will be subjected to a high radiation environment due to the incoming cosmic rays, comprising protons, α particles, electrons, γ rays, etc., once they escape the Earth’s atmosphere. The ion irradiation of CF-PEEK is accompanied by radiation-induced effects, which drastically change the structure and properties of irradiated material. Since the resistance of CF-PEEK to radiation damage has not been studied extensively, this study aims to understand the effect of high-energy He 2+ ions on the microstructure and properties of CF-PEEK composites manufactured using automated fibre placement (AFP) under different processing conditions. The samples have been radiated with 5 MeV He 2+ ions using an energy degrader wheel to create a layer with relatively uniform damage. Then, were characterized using optical and scanning electron microscopy and their hardness was evaluated using nanoindentation. It was observed that, irradiation increases the hardness of the fibres in all cases. Also, fibre orientation affects the hardness in a statistically significant manner in both unirradiated and irradiated conditions.

Abstract Lightweight, strong, and radiation-tolerant materials are essential for advanced nuclear systems and aerospace applications. However, the current research for these materials mainly focuses on a single design strategy to improve their irradiation resistance via empirical trial-and-error learning. In this study, a novel NbVTaSi refractory eutectic high-entropy alloy was target designed via high-throughput thermodynamic calculations. To exploit the unique eutectic structures of NbVTa and β-Nb5Si3 phases, a new synergistic mechanism for irradiation resistance via balancing inhibition effects on the generation of He-induced lattice defects and the formation and growth of He bubbles was proposed according to the experimental findings and density functional theory calculations. The alloy possesses lightweight (7.4 g/cm3), high yield strengths at room temperature (2.60 GPa) and 850 ℃ (1.84 GPa), and outstanding He-irradiation resistance, superior to other reported radiation-resistant alloys. This study sheds light on the development of future radiation-tolerant materials for advanced nuclear systems and aerospace applications.

A density functional theory (DFT) study was employed to investigate the mechanical property, thermal conductivity, Debye temperature, electronic structure and defect chemistry of (Gd1-xSmx)2Zr2O7. All the (Gd1-xSmx)2Zr2O7 compounds exhibit an excellent structural and mechanical stability (Gd0.25Sm0.75)2Zr2O7 has the lowest Young's modulus of 213.7 GPa, the largest Possion's ratio of 0.292, the lowest Debye temperature of 491.8 K and the lowest thermal conductivity. The calculated thermal conductivities of (Gd1-xSmx)2Zr2O7 are 1.17-1.21 W/(m center dot K) by the Clark's model and 1.32-1.36 W/ (m center dot K) by the Cahall's model, respectively. The formation energies of O vacancies at 48f site are negative, which increase with the Sm content, however, the formation energies of O vacancies at 8b site are almost invariable. In addition, Sm partly occupying the Gd-site reduces distinctly the formation energies of defects such as A-site vacancies, cation antisite defects, anion Frenkel pairs of oxygen at 8b site and cation interstitials, which suggests that Sm-doped Gd2Zr2O7, especially equimolar GdSmZr2O7, has a better irradiation tolerance. After the 16 MeV Ta-ion irradiation at a fluence of 1 x 1014 or 2 x 1014 ions/cm2, the crystal structure of GdSmZr2O7 transforms from pyrochlore to a defect fluorite without obvious amor-phous phase.(c) 2022 Chinese Society of Rare Earths. Published by Elsevier B.V. All rights reserved.