Dry reprocessing technology has advantages of irradiation resistance, proliferation resistance and simplified waste treatment, thereby rendering it a preferred technology for reprocessing of spent fuel of advanced fast reactors. Molten salt electrolytic refining serves as the core technology of dry reprocessing, primarily capitalizing on the difference in redox potential between actinides such as uranium and plutonium and other fissionable elements within a molten salt system. This technology facilitates separation and recovery of actinides. During the electrolytic refining process, lanthanide elements and fission elements, such as Sr and Cs, tend to accumulate within the molten salt, which can change the physicochemical properties of molten salt, thus seriously affecting the efficiency of electrolytic refining. In addition, fission products such as ⁹⁰Sr and ¹³⁷Cs are water-soluble and long-lived nuclides, posing significant environmental hazards if inadequately managed. Therefore, effective purification of fission elements such as Sr and Cs from molten salt is imperative, not only to improve the efficiency of dry reprocessing via molten salt electrolysis, but also as a crucial strategy to reduce the discharge of radioactive waste. This paper summarizes the current research status of Sr and Cs removal methods in molten salts, comparatively analyzes the separation principles and separation effects of different methods, such as electrolysis, crystallization, decompression distillation, precipitation, and ion exchange. Furthermore, it explores prospective direction of development and potential applicability of various material systems.

Lead-bismuth eutectic (LBE), with its high boiling point, excellent thermal conductivity, and favorable safety properties, has become the core coolant for accelerator-driven advanced nuclear energy systems (ADANES) and lead-cooled fast reactors (LFRs). However, the coupled issues of oxidation corrosion, dissolution corrosion, and erosion corrosion induced by its high-temperature environment severely threaten the service life of structural materials, and hinder the engineering implementation of advanced nuclear technology. Surface coating technology, which enhances corrosion resistance while preserving the inherent properties of the substrate, has emerged as a key strategy to mitigate LBE corrosion. This paper provides a systematic review of research progress on corrosion-resistant coatings for nuclear applications against LBE corrosion. Starting from corrosion mechanisms, the influences of synergistic factors, including dissolved oxygen, temperature, flow velocity, and irradiation, on corrosion behavior are elucidated. Coatings are classified into three major systems, i.e. metallic, ceramic, and composite, and the corrosion resistance mechanisms, performance advantages, and failure behaviors of FeCrAl(Y), high-entropy alloys, Al₂O₃, MAX phases, and gradient composite coatings are analyzed. Research indicates that FeCrAl(Y) coatings form a continuous Al₂O₃ barrier layer through “matrix-oxide film” synergistic effect, with corrosion resistance and mechanical properties correlated to Cr and Al content. High-entropy alloy coatings suppress inward diffusion of corrosion species through lattice distortion and multi-component synergistic oxidation, yet they are susceptible to like high-temperature phase decomposition and irradiation embrittlement. Thermodynamically stable ceramic coatings like Al₂O₃ provide effective substrate protection, yet tend to fail in high-temperature environments due to amorphous crystallization, interface mismatch, and lack of self-healing. Composite-structured coatings with a gradient architecture of “metal transition layers + ceramic functional layers” present a promising approach for integrating high adhesion, high toughness, and high barrier properties. Future efforts should focus on coupled regulation of environment-composition-process, multi-factor service evaluation, and lifetime prediction models to deliver long-term reliable protection solutions for advanced nuclear energy systems.

Silicon carbide (SiC) is a promising material for nuclear reactor structures due to its excellent radiation resistance and high-temperature performance. The behavior of irradiation damage and the mechanisms of high-temperature recovery in SiC directly affect its service performance and longevity in nuclear environments. This study investigated effects of neutron irradiation on properties of 6H-SiC, with a particular focus on high-temperature recovery mechanisms of irradiation-induced defects. Specifically, defect evolution and thermodynamic responses in nitrogen-doped (N_D≈3.0×10¹⁹ cm^-3) 6H-SiC subjected to neutron irradiation at about 150 ℃ and a fluence of 2.58×10²⁰ n/cm² followed by isochronal annealing were examined. Integrated techniques and first-principles calculations were employed to comprehensively analyze its structural and property evolution. The key findings were as follows. (1) Significant lattice swelling was observed during the irradiation, with a swelling rate of 0.416% along the a-axis, 0.430% along the c-axis, and 1.310% in the unit cell volume, while all maintaining integrity of the single-crystalline structure. (2) A 14.7% increase in specific heat capacity was recorded, with 375.4 J/g of stored irradiation energy being released during heating from 100 ℃ to 500 ℃. (3) A four-stage defect recovery kinetic model was proposed based on the recovery of lattice parameters and the evolution of Raman spectra: Stage I (room temperature (RT)-600 ℃), primarily dominated by close-range recombination of carbon Frenkel pairs driven by migration energy (E_a) of 0.14 eV; Stage II (600-850 ℃), recombination of silicon Frenkel pairs and migration of carbon interstitials (E_a=0.26 eV); Stage III (850-1200 ℃), lattice reconstruction (E_a=0.65 eV); Stage IV (1200-1500 ℃), long-range diffusion of carbon vacancies (V_C) and dissociation of N_CV_Si complexes (E_a=1.50 eV). (4) The presence of nitrogen-stabilized N_CV_Si defect configurations was confirmed by a characteristic emission peak at 826 nm (634 cm^-1 Raman shift) when excited with 785 nm light. This study quantitatively reveals the defect recovery pathways and migration energies in neutron-irradiated 6H-SiC, providing a critical foundation for evaluating radiation damage, predicting performance, and optimizing annealing processes in nuclear-grade SiC materials.

Mining process of natural uranium ore generates uranium-containing wastewater, while removal of uranium(VI) from such wastewater has emerged as a critical challenge, requiring urgent resolution in the nuclear industry. Guided by the principle of "from uranium mines, back to uranium mines," this study selected calcium orthovanadate (Ca₃(VO₄)₂) as an adsorbent for U(VI) removal. Adsorption performance of Ca₃(VO₄)₂ powder under varying conditions and its underlying mechanism were investigated. Results demonstrated that under optimal conditions (pH 6, adsorption for 2 h, adsorbent dosage at 0.1 g·L^-1, initial U(VI) mass concentration at 120 mg·L^-1, temperature at 308 K), Ca₃(VO₄)₂ powder exhibited a high adsorption capacity (1179.92 mg·g^-1) and removal efficiency (98.33%) for U(VI). Removal mechanism was attributed to dissolution and mineralization processes, forming metatyuyamunite (Ca(UO₂)₂(VO₄)₂·3H₂O) on the powder surface after adsorption. Even in the environment of six coexisting ions (Zn²⁺, Cr³⁺, Cu²⁺, Ni²⁺, Co²⁺, and Ba²⁺), Ca₃(VO₄)₂ maintained high adsorption performance, reducing U(VI) mass concentration from 121.49 to 0.1 mg·L^-1, which is below the limit specified by the national discharge standard (GB 23727-2020). These findings highlight Ca₃(VO₄)₂ as a promising adsorbent for efficient treatment of U(VI)-containing wastewater.

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.

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.

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.