8% (molar fraction) Y₂O₃ stabilized ZrO₂ (8YSZ) ceramics have important applications in fuel cells, thermal barrier coatings, as well as thermal insulation due to their excellent oxygen ionic conductivity and low thermal conductivity. However, their corrosion resistance to water and their behaviors as thermal insulation or structural material in pressurized water reactors during accidents are not fully understood. This study systematically examined the mass, crystal phase, microstructure, mechanical properties, and solution composition of 8YSZ ceramics over time in a dynamic water environment at 350 ℃/17.4 MPa with 0.3 μg/L dissolved oxygen, aiming to simulate a pressurized water reactor environment. It is found that the mass of 8YSZ ceramics increases firstly and then decreases with corrosion duration time. The mass change is influenced by the surface roughness. The weight gain is attributed to the formation of Zr-OH and Y-OH clusters by the entry of water molecules into the ceramics, whereas the weight loss is caused by the metal cations leaching and the dissolution of grains. Phase analysis demonstrates that the cubic 8YSZ after corrosion does not undergo any phase transformation towards tetragonal or monoclinic phases, which is different from the degradation mechanism of tetragonal or partially stabilized zirconia. Changes in surface and cross-section morphology indicate that water molecules enter the interior of the ceramics along defects or microcracks, producing grain boundary damage and changing the fracture mode in the corrosion-affected region from transgranular to intergranular fracture. Compressive and flexural strengths of this ceramics after corrosion do not change significantly, while the Vicker’s hardness decreases slightly, which are related to the formation of pits in the surface layer. As a consequence, depth of the corrosion pit after 1050 h is only 30.8 μm, and the mass change rate of per unit surface area is -0.108×10^-3 mg∙cm^-2∙h^-1, consolidating excellent water corrosion resistance of 8YSZ ceramic. Therefore, 8YSZ ceramics are promising for thermal insulation or structural materials in pressurized water reactors.

Fully ceramic micro-encapsulated (FCM) fuel has become the focus of nuclear energy research because of its good inherent safety. In order to overcome the difficulty of SiC matrix densification, this study focused on the sintering behavior of simulating core FCM fuel via hot oscillatory pressing (HOP), taking the advantages of HOP to accelerate mass transfer and reduce sintering temperature. The influence of oscillatory sintering temperature, oscillatory time and oscillatory pressure on the matrix densification behavior was studied, and the results were compared with those of hot pressing (HP). The results indicate that the oscillatory sintering temperature, holding time, and median pressure have important effects on matrix densification, while the amplitude of oscillatory pressure has little effect. Compared with HP, the density of the samples is increased by HOP, and the density of sample sintered at 1850 ℃ via HOP is 99.99%. The grain size of the samples via HOP is smaller, and the grain size of the sample sintered via HOP at 1850 ℃ is (284±4) nm, which is ~27% less than that of the sample sintered via HP at the same temperature. The hardness of the samples sintered via HOP is higher, and the hardness of the sample sintered via HOP at 1850 ℃ is (26.7±0.4) GPa. When the density of sample is 90%, the stress exponent n=1 and activation energy Q=430 kJ/mol are obtained by using the modified constitutive equation of HP. The dominant mechanism of densification is grain boundary sliding, which is accommodated by the grain boundary diffusion.

Neutron detection technology is widely used in homeland security, nuclear material security detection, and high energy physics, etc. Due to the shortage of ³He resources, it is urgent to develop a novel scintillator that can discriminate neutron and gamma. The Cs₂LaLiBr₆:Ce (CLLB:Ce) crystal has good neutron/gamma discrimination capacity, excellent energy resolution and high light yield, but its neutron/gamma discrimination performance needs further improvement. Here, the CLLB:Ce crystals co-doped with Zr⁴⁺ were grown successfully by the vertical Bridgman method. The results of different characterization methods prove that the Zr⁴⁺ was successfully doped into the matrix and did not effect on the structure of host. Meanwhile, no new luminescence center was generated after Zr⁴⁺ doping. The UV decay time is about 27 ns, presenting a fast fluorescence decay. Figure of merit (FOM) of CLLB:Ce crystal is enhanced from 1.2 to 1.5 by co-doping Zr⁴⁺, which means that the neutron/gamma discrimination performance of CLLB:Ce crystals is improved. Combined with the thermal stability and scintillation decay time, relationship between decay time and FOM was also analyzed. The co-doping of Zr⁴⁺ can inhibit shallow electron trap and V_K centers, reduce electron trapping-detrapping process, and greatly increase the probability of Ce³⁺ direct capturing electron, which results in a shorter decay time. Data from this study indicate that the CLLB:Ce crystals exhibit a huge application prospect in the field of neutron/gamma detection.

Molten salt electrolysis is the key technology for dry reprocessing of spent fuel in the nuclear energy industry. High-temperature molten salt can cause severe corrosion to crucible materials used for spent fuel, so the selection of the crucible material with good resistance to high temperature and corrosion is crucial for the development of the dry reprocessing method. Si₃N₄ is considered as a promising candidate for the crucible used in dry reprocessing, primarily due to its excellent high-temperature thermal and mechanical properties. However, its resistance to high-temperature molten salts and water vapor has not been fully investigated. In this work, the corrosion behavior of Si₃N₄ in LiCl-KCl and NaCl-2CsCl molten salt under Ar atmosphere and water vapor (5%H₂O-10%O₂-85%Ar) was investigated. The results show that in argon atmosphere, Si₃N₄ undergoes slight grain boundary corrosion in LiCl-KCl molten salt, while NaCl-2CsCl molten salt presents weak corrosion on Si₃N₄. In 5%H₂O-10%O₂-85%Ar water vapor environment, LiCl-KCl molten salt prefers to attack the grain boundary phase. Si₃N₄ shows serious corrosion degradation in the NaCl-2CsCl molten salt compared with the corrosion level in argon atmosphere. The water vapor environment significantly promotes the corrosion of Si₃N₄ in the molten salt environment, while the grain boundary phase is the most susceptible site for the corrosion of Si₃N₄. In addition, no direct correlation is found between the wettability and corrosion resistance of LiCl-KCl and NaCl-2CsCl molten salts. Results of this work elucidate the mechanism of high-temperature molten salt-water vapor-induced degradation of Si₃N₄, offering guidelines for the selection of crucibles in the dry reprocessing of spent fuel.

Perovskite-structured praseodymium aluminum oxide (PrAlO₃) exhibits high stability which has a site that can be doped with other rare earth ions, enabling it a promising new neutron-absorbing material matrix. However, the current research on PrAlO₃ mainly focuses on the preparation methods of single crystal materials and their optical and magnetic property. Here, we firstly prepared a high-density perovskite phase PrAlO₃ ceramics by solid-phase reaction synthesis using tetraethyl orthosilicate (TEOS) as a liquid phase sintering aid, and then studied its microstructure and thermal property by XRD, SEM, push-rod technique, and laser flash method. The results showed that, by pre-synthesizing PrAlO₃ powder at 1200 ℃ and adding 0.4%-1.0%(in mass) TEOS as a liquid phase sintering aid, PrAlO₃ ceramic with a relative density higher than 99% could be obtained at around 1500 ℃, while the relative density of the product without sintering aids was only 96%. The thermal conductivity of PrAlO₃ ceramic at a subcritical temperature of 360 ℃ was 4.99 W·m^-1·K^-1, superior to those of Dy₂TiO₅ and GdAlO₃ ceramics, and its linear thermal expansion coefficient from room temperature to 800 ℃ was only 10.2×10^-6 K^-1. Moreover, the bending strength and Vickers hardness of PrAlO₃ ceramics reached 95.55 MPa and 7.95 GPa, respectively, and the fluorescence spectrum exhibited characteristic emission peaks of Pr³⁺. This study shows that high-density perovskite phase PrAlO₃ ceramics can be prepared by a convenient method with good thermophysical property and mechanical property. They exhibit good application prospects as a rare earth-based neutron-absorbing nuclear material.

MAX/MAB phases are a series of non-van der Waals ternary layered ceramic materials with a hexagonal structure, rich in elemental composition and crystal structure, and embody physical properties of both ceramics and metals. They exhibit great potential for applications in extreme environments such as high temperature, strong corrosion, and irradiation. In recent years, two-dimensional (2D) materials derived from the MAX/MAB phase (MXene and MBene) have attracted enormous interest in the fields of materials physics and materials chemistry and become a new 2D van der Waals material after graphene and transition metal dichalcogenides. Therefore, structural modulation of MAX/MAB phase materials is essential for understanding the intrinsic properties of this broad class of layered ceramics and for investigating the functional properties of their derived structures. In this paper, we summarize new developments in MAX/MAB phases in recent years in terms of structural modulation, theoretical calculation, and fundamental application research and provide an outlook on the key challenges and prospects for the future development of these layered materials.

In contrast to conventional solid-phase sintering, molten salt method can provide a fast mass transfer and nucleation process at lower temperatures which has potential to synthesize the ceramic solid solution for immobilization of high-level nuclear waste (HLW). In this work, Nd-doped zircon (ZrSiO₄) ceramics (Zr_1-xNd_xSiO_4-x/2 (0≤x≤0.1)) were prepared by the molten salt synthesis(MSS) at different sintering temperatures (1100, 1200, 1300, 1400, 1500 ℃) for different sintering time (3, 6, 9, 12, and 15 h). Chemical stability of Nd-doped zircon ceramics in simulated geological disposal environment was studied by static leaching test (PCT). Zr_1-xNd_xSiO_4-x/2 was synthesized by the molten salt method under the optimum molar ratio of molten salt to oxide at 10:1, sintering temperature at 1200 ℃ and sintering time of 6 h with the solid solution of Nd in ZrSiO₄ being increased to 8% (in mol). The MSS can reduce the synthetic temperature, shorten the sintering time and save the solid solution. The immobilizing mechanism of ZrSiO₄ ceramics for trivalent actinide nuclides is lattice immobilizing. Experimental results show that the normalized leaching rate (LR_Nd) of Nd is as low as ~10^-5 g·m^-2·d^-1. ZrSiO₄ ceramics have no phase evolution before and after leaching, suggesting good structural stability. TLeaching model consolidates that Nd leaching is due to dissolution of the ceramic surface layer. Data from this study show that MSS is a promising method to synthesize ceramics solid solution.

High energy particle bombardment of silicon carbide can lead to the accumulation of defects and lattice disorder, which can negatively affect physical property and reduce lifetime of SiC devices. Thus, it is essential to systematically study the damage of SiC in different radiation environment. Herein, 6H-SiC was irradiated by neutrons at the fluence of 5.74×10¹⁸, 1.74×10¹⁹, 2.58×10²⁰ and 1.27×10²¹ n/cm², and then annealed. Changes in lattice parameters from post-irradiation isochronal annealing for 30 min in the range of 500-1650 ℃ were measured using X-ray single crystal diffraction. The results showed that the lattice swelling and recovery behavior were isotropic. Based on the swelling data, it was concluded that the neutron irradiation-induced defects in 6H-SiC were primarity point defects. Both intrinsic and irradiation defects can introduce defect energy levels, which were mainly caused by vacancies and led to the absorption band edge redshift and band gap narrowing of SiC. The defect energy levels of these vacancies and vacancy-associated defects were determined by absorption spectra, luminescence spectra and Raman spectra. Experiments and first principles calculation showed that the silicon vacancies introduced defect levels above the valence band, while the carbon vacancies introduced levels below the conduction band. The infrared absorption at 1382 nm and 1685 nm and the emission at 550 nm of unirradiated 6H-SiC were mainly due to the intrinsic carbon vacancies. The luminescence of post-irradiated SiC at 415, 440 and 470 nm was mainly due to the silicon vacancy produced by irradiation and its related defect configuration. All above data revealed the luminescence mechanism of SiC based on the charge state and the defect energy level distribution.

Silicon carbide fiber reinforced silicon carbide (SiC_f/SiC) composites have become the preferred candidate for structural applications in advanced nuclear energy systems, because of their low neutron toxicity, neutron irradiation tolerance and high-temperature oxidation resistance. In recent years, both academia and industry either domestic or abroad have carried out a lot of researches on SiC_f/SiC composites for nuclear application, and numerous important achievements have been made. This paper summarized and analysed some critical directions of SiC_f/SiC composites for nuclear applications, including nuclear-grade SiC fibers, fibre/matrix interfaces, composite processing, modeling and simulation, corrosion behavior and surface protection, joining technology, as well as radiation damage. The key issues and potential solutions of SiC_f/SiC composites for nuclear applications have been pointed out in account to the requirements, anticipating to be beneficial to promoting further researches and final applications.