Vitrification is a widely used approach for the immobilization of high-level liquid waste. In this work, the melting behavior and phase evolution of simulated HLW calcine in a borosilicate glass matrix were investigated as a function of waste loading and melting temperature. The phase assemblage, microstructure, and chemical states were examined using SEM, XRD, FT-IR, Raman spectroscopy, and XPS.The results indicate that the interaction between the calcine and the glass matrix initiates at approximately 700 degrees C, accompanied by the appearance of diffraction features assigned to a Ca-Mo-O phase, consistent with CaMoO4. The melting reaction becomes more pronounced at around 900 degrees C. At 1100 degrees C, the majority of the calcine is incorporated into the glass matrix, while weak diffraction signals attributed to fluorite-related rare-earth-cerium oxide solid solutions, denoted as MxCe1-XO2 (M=Nd/Sm/La/Pr), remain detectable. Waste loading exerts a strong influence on the melting behavior and phase assemblage. Increasing the waste loading from 24 wt% to 32 wt% results in an increase in the temperature required for effective melting, and samples with 32 wt% waste loading retain detectable fluorite-related oxide solid-solution phases even at 1100 degrees C.These results clarify the temperature-dependent melting behavior and phase evolution of HLW calcine in borosilicate glass and demonstrate the influence of calcine content on phase transformation temperatures during vitrification.

Efficient and selective uranium extraction from wastewater remains a significant challenge, primarily because conventional adsorbents exhibit low affinity for UO22+, while most catalytic centers are either neutral or positively charged, creating a fundamental charge incompatibility with the target ions. To overcome this, we adopted an integrated adsorption-electrocatalysis strategy using a freestanding membrane electrode (ZMOM), fabricated via electrostatic self-assembly of Ti3C2Tx MXene and zinc molybdate (ZMO). The key innovation lies in employing molybdate ions as bifunctional units that combine a negatively charged surface for electrostatic attraction of UO22+ with oxygen-bridged configurations that serves as integrated catalytic sites. This synergy enables continuous capture and molybdenum-mediated redox conversion (U6+ ↔ U5+), leading to the precipitation of insoluble Na2O(UO3·H2O)x for easy recovery. Under the square-wave exchange (SWE) method, the optimized ZMOM-2 electrode achieves a high uranium extraction capacity of 1641.78 mg/g at -5 V from a 100 mg/L solution and retains 98.78% of its capacity over five consecutive cycles. This work establishes a new design principle that integrates adsorption sites and anionic catalytic centers into a single material, offering an efficient and robust route for uranium resource recovery.

Vitrification is worldwide recognized as the most reliable technology for high level liquid waste(HLLW) immobilization. At present, the mainstream technologies for vitrification of high level liquid waste(HLLW) worldwide include joule-heated ceramic melter vitrification and two-step cold crucible induction melter(CCIM) vitrification. In the two-step cold crucible vitrification process, HLLW undergoes denitration and calcination sequentially in a rotary calciner to form calcined waste. The calcine is then fed into a cold crucible induction melter, where it is melted together with a glass-forming batch. Upon cooling, a homogeneous waste form is produced, achieving safe immobilization of radioactive waste. This study is based on the two-step cold crucible induction melter vitrification technology. Previous studies regarding glass formulation and process optimization have mainly focused on the direct treatment of high-level radioactive liquid waste. However, to meet practical engineering operation requirements, slag flushing water co-generated during reprocessing must also be co-treated. According to source-term characterization, the main component of slag flushing water is metallic zirconium scraps(accounting for more than 90%(mass fraction, the same below)), accompanied by minor insoluble radionuclides including U, Pu, Ru, and Tc. Therefore, during the co-treatment of HLLW and slag flushing water, the effects of metallic zirconium scraps on the vitrification process and glass formulation must be carefully evaluated. In the two-step cold crucible vitrification process, metallic zirconium scraps are converted into zirconia(ZrO2) during the first-stage rotary calcination. From the perspective of glass formulation design, the influence of slag flushing water is thus dominated by the effects of ZrO2 on the structure and performance of the final glass waste form. In this study, glasses containing simulated HLLW and different amounts of ZrO2 were prepared. The results show that the glass wasteform structure is able to accommodate up to 6% ZrO2 content. With the increase of ZrO2 doping content in the borosilicate glass matrix, the glass density rises obviously while the molar volume decreases; meanwhile, both the chemical durability and high-temperature viscosity of the glass are enhanced. However, when 2% RuO2 coexists in glass, the glass density and molar volume slightly change with ZrO2 content while the glass viscosity largely increases. Moreover, the nuclear magnetic resonance(NMR) 29Si and 11B results suggest that ZrO2 will depolymerize the [SiO4] connections in network backbone, whereas increase the ratio of [BO4]/[BO3], thereby increasing the stability of glass network.

The MXene-induced interfacial Schottky barrier creates a built-in electric field that drives spontaneous electron transfer from Mn 0.3 Cd 0.7 S to Ti 3 C 2 T x , achieving 100% U( vi )-to-U( iv ) photoreduction within 30 min through ˙O 2 − -mediated chain reactions.

A one-step etching-functionalization strategy that preserves MXene structural integrity while grafting active sulfonic groups is reported herein. The self-supporting TCS membrane achieves 3576.3 mg g-1 uranium extraction and retains 86% of its initial capacity after ten cycles, offering a scalable route for efficient recovery.

High-level liquid waste (HLLW) is characterized by high radioactivity, long half-life and high biological toxicity; therefore, its safe management remains a major issue for nuclear fuel-cycle applications. Glass-ceramic waste forms combine the compositional flexibility and homogeneity of glass with the selective incorporation capacity of crystalline phases, and are promising matrices for immobilizing heavy metals and rare-earth elements in HLLW. In this work, simulated high-level waste was immobilized in sodium-aluminum-iron phosphate glass-ceramics. X-ray diffraction, scanning electron microscopy with energy-dispersive spectroscopy, Fourier-transform infrared spectroscopy, Raman spectroscopy and X-ray photoelectron spectroscopy were used to examine phase formation, microstructure, network evolution and chemical states as a function of waste loading. The results show that the waste forms remain predominantly amorphous at waste loadings of 8 wt.% and below. At 14 wt.%, the phosphate network is significantly modified and monazite-type phosphate crystalline phases, including Ce(PO4), Ca0.5Ce2(PO4)3 and Sr0.5Ce2(PO4)3, precipitate extensively. At waste loadings of 16 wt.% and above, the diffraction peaks of the monazite-type phosphate phases become stronger with increasing waste content, indicating an increased degree of glass-ceramic crystallization. These results provide a structural basis for the design of phosphate glass-ceramic matrices for HLLW immobilization.

As a pivotal technology for high-level liquid waste (HLW) treatment, vitrification plays a crucial role in ensuring the sustainable development of nuclear energy. The two-step vitrification process, which involves initial calcination of HLW into solid calcine followed by melting with base glass, has demonstrated significant improvements in immobilization efficiency. This study investigated the capacity of simulated HLW calcine to be incorporated within a borosilicate glass matrix.Through comprehensive characterization techniques including field emission scanning electron microscopy coupled with energy dispersive spectroscopy (SEM-EDS), X-ray diffraction (XRD), Fourier-transform infrared spectroscopy (FTIR), and Raman spectroscopy, the effects of calcine content (18-32 wt%) on the microstructure, chemical composition, phase distribution, and network structure of the vitrified matrixs were examined. Homogeneous glass matrices are achieved at 18-26 wt% calcine content, while phase separation and crystalline precipitation occur at higher concentrations. However, when the content reaches 28 wt%, the initial precipitation of granular CaMoO4 crystals occurs, maintaining relatively uniform distribution. Further increasing the content to 30 wt% leads to the formation of acicular CaMoO4 phases accompanied by minor ZrxCe1-xO2 particle precipitation. Notably, at 32 wt% calcined product content, a distinct phase separation layer emerges on the surface of the vitrified matrix, with a substantial increase in ZrxCe1-xO2 precipitates. These research findings have clarified the incorporation capacity limits of simulated HLW calcine in borosilicate glass matrix, providing valuable theoretical guidance for enhancing waste loading in vitrified forms during the vitrification process.

The geological disposal of high-level liquid waste (HLW) generated from spent fuel reprocessing after vitrification is widely recognized as a viable solution with manageable impacts on the natural environment and human health. A critical factor influencing the long-term safety of this process is the leach resistance of the glass. In this study, calcined products of simulated HLW and a borosilicate glass matrix were utilized as raw materials. Glass samples with waste loadings of 18wt%, 21wt%, and 24wt% were prepared using the melting method, and their leaching behaviors were systematically assessed through the MCC-1 static leaching method. By analyzing the leaching kinetics of various elements, observing changes in surface morphology, monitoring the thickness of the corrosion layer, and examining the elemental distribution characteristics of the waste form during the leaching process, we elucidated the microscopic mechanisms underlying leaching. The experimental results indicate that as waste loading increases, the total weight loss per unit area of the waste form correspondingly rises. Through the MCC-1 standard leaching process, the standardized leaching rates of the important elements in the glass were all lower than 0.45 g·m−2·d−1, indicating good overall stability. In terms of leaching behavior, elements such as B, Na, Si, and Ca in the glass matrix exhibit relatively high leaching rates, while Mo and Cs in the simulated waste components are also prone to leaching. Conversely, elements like Zr and La demonstrate extremely low leaching rates, indicating excellent leach resistance. A 120-day long-term leaching experiment on the waste form with a waste loading of 21 wt% further reveals that during the initial leaching phase (1-28 days), the surface of the waste material gradually forms a corrosion layer from a flat state. After 60 days, the thickness of the corrosion layer stabilizes, resulting in the formation of a dense reaction layer. This layer primarily consists of a gel formed by surface hydrolysis, which effectively obstructs direct contact between the leaching agent and the bulk of the waste form, significantly reducing subsequent rates of ion exchange and hydrolytic desilication, thereby ensuring the chemical stability of the waste form in a long-term disposal environment.

Zirconium is a key fission product in high-level liquid waste, originating from the neutron-induced fission of uranium and plutonium in nuclear fuel, and its content in the waste stream can be substantial depending on burnup and cooling time. Its solubility in borosilicate glass directly governs the achievable waste loading and the long-term chemical durability of the solidified waste form, because excess ZrO2 tends to precipitate as refractory phases that compromise the glass network integrity. In this study, a systematic experimental investigation was conducted to determine the solubility thresholds of ZrO2 in a typical borosilicate glass system by precisely controlling two variables: the total simulated waste content(representing a multi-component oxide mixture) and the added ZrO2 amount. Two distinct addition modes were examined in parallel. The first mode, “tolerance capacity”, involved substituting ZrO2 for an equivalent mass of the simulated waste mixture, thereby assessing the maximum tolerable ZrO2 level while keeping the total waste loading constant. The second mode, “co-solubility”, involved adding ZrO2 simultaneously with the full simulated waste composition to evaluate the mutual solubility limits under realistic multi-component conditions, where various cations compete for network-modifying sites. Waste loadings were set at 18%, 20%, and 22%(mass fraction), and ZrO2 content was varied across a wide range to bracket the expected thresholds. Characterization of the glass waste forms was performed using scanning electron microscopy(SEM) for microstructure observation, energy-dispersive spectroscopy(EDS) for elemental mapping to detect Zr-rich segregations, and X-ray diffraction(XRD) for unambiguous identification of crystalline phases. The results reveal a clear dependence of the ZrO2 solubility limit on both waste loading and addition mode. Under tolerance capacity, at waste loadings of 18% and 20%, the solubility threshold remains stable at 6%-8%(mass fraction) ZrO2, indicating sufficient free volume and modifier cations to incorporate Zr into the silicate network. However, when the loading increases to 22%, the threshold drops significantly to 4%-6%, because the available network modifier sites become increasingly occupied by other waste components such as rare earths and transition metals, reducing the accommodation capacity for Zr. Under co-solubility, with simulated waste content ranging from 18% to 22%(mass fraction), the threshold is consistently 4%-6%, indicating that simultaneous presence of all waste constituents imposes a more restrictive limit due to competitive cation interactions and the saturation of the glass network. Importantly, exceeding the respective threshold leads to pronounced phase separation, with crystallization of ZrSiO4 and residual ZrO2 phases, as confirmed by XRD. This transformation from a homogeneous glassy state to a heterogeneous composite not only impairs the chemical durability—increasing the leaching rate of radionuclides—but also reduces the mechanical strength and thermal stability of the waste form. The present work establishes quantitative solubility boundaries for ZrO2 under realistic conditions, provides mechanistic insights into the competition between glass network incorporation and phase precipitation, and offers critical data for optimizing ZrO2 loading and adjusting glass formulations in practical vitrification processes. These findings are of significant engineering value for improving waste volume reduction, minimizing secondary waste, and ensuring long-term safety in nuclear waste management, thereby supporting the design of more robust and efficient glass waste forms for geological disposal.

Inspired by cellular slime molds, a biomimetic self-supporting MXene membrane is fabricated. This design integrates covalent cross-linking for stability with phosphate groups for specific uranium capture. The membrane achieves a high extraction capacity of 3776.25 mg g-1 under an applied voltage, demonstrating a promising strategy for advanced nuclear wastewater treatment.

Effective U(VI) separation from uranium wastewater is crucial for nuclear energy's sustainable growth. A magnetic, cauliflower-shaped magnesium-iron bimetallic oxide loaded biochar (MFBC) was fabricated via a green hydrothermal co-pyrolysis. Its U(VI) adsorption process was pH-dependent and fitting Langmuir–Freundlich and pseudo-second-order models, the theoretical maximum adsorption capacity of U(VI) by MFBC peaked at 1354.8 mg·g−1 (pH 5.0, 30 min). After five regenerations, 90.3

Nitrogen-doping of graphene oxide can significantly enhance its electrochemical and electrosorption performances for uranium(VI). The conventional methods of N-doping such as hydrothermal and pyrolytic processes invariably cause a significant loss of surface groups, which are not conducive to further functionalization of graphene oxide with other polymers by means of chemical bonds. Herein, a novel N-doped graphene oxide was first prepared using graphite by a precursor reconfiguration strategy, and then used as the precursor to construct a self-supported N-doped graphene oxide aerogel (PNGOA) by freeze drying technique. The prepared PNGOA exhibited larger specific surface area, better hydrophilicity, more structural defects, superior electrochemical and electrosorption performances. Next, the hydrophilic polyethyleneimine (PEI) was integrated into PNGOA to synthesis the N-doped graphene oxide aerogel/polyethyleneimine (PNGOA/PEI) electrode in order to further enhance the electrosorption of uranium(VI). The assessed theoretical maximum uranium(VI) electrosorption capacity by the PNGOA/PEI electrode was determined to be 680.89 mg/g, which was twice that of graphene oxide aerogel/polyethyleneimine (GOA/PEI) electrode. PNGOA/PEI aligned more closely with pseudo-second order kinetic and Langmuir model. The effective electrosorption of U(VI) by the PNGOA/PEI electrode benefits from the improvement of electrochemical properties of graphene oxide due to its N-doping by a precursor reconfiguration strategy, and the synergistic interactions between the -NH2, -CONH-, and -OH groups from PEI with uranium(VI) on the PNGOA/PEI electrode. This work greatly inspired the potential of graphene oxide aerogel via an innovative method for effectively eliminating uranium(VI) from wastewater contaminated with uranium.

Electrosorption technique is paid more attention in recovery of uranium(VI) due to its low energy consumption, green process and easy regeneration, but still is hindered by the limited accessible active sites on the electrode material, poor surface wettability and inherent ion exclusion effects. Herein, novel there dimensional porous polyethyleneimine/graphene oxide aerogel (PEI/GOA) electrode was constructed by incorporating polyethyleneimine into graphene oxide and using chitosan as binder for uranium(VI) electrosorption. The synthesized PEI/GOA had a stable three-dimensional porous structure, high specific capacitance (54.06F/G) and good charge -discharge stability. The electrosorption process of uranium(VI) by PEI/GOA electrode was affected by applied voltage and pH, fitted by the pesuo-second-order kinetics and Langmuir model. The maximum theoretical electrosorption capacity of uranium(VI) by PEI/GOA electrode at - 1.2 V was 419.71 mg/g, which was 2.2 times higher than that of pure graphene oxide electrode. PEI/GOA electrode exhibited less than 20 % loss after five cycles using 0.1 M NaHCO 3 as the eluent. The efficient U(VI) electrosorption by PEI/GOA electrode was not only attributed to the formation of the electric double layer, but also due to the interaction and synergy of the -NH 2 , -CONH- and -OH on the PEI/GOA electrode. This research offers a new approach for the efficient removal of uranium (VI) from uranium -containing wastewater.

The electrocatalytic role of Ti 3 C 2 T x MXene in U(VI) immobilization has remained largely unexplored. Herein, a binder‐free electrode (TiMX/CNT‐COOH) is designed and its exceptional performance in electrochemically U(VI) extraction under square‐wave exchange (SWE) is demonstrated. The incorporation of carboxylated carbon nanotubes (CNT‐COOH) as a rigid spacer not only enhances structural disorder but also exposes abundant undercoordinated Ti edge sites and induces bond stretching (Ti─O and O─H), further boosting intrinsic catalytic activity. This synergistic effect interaction lowers the energy barrier of the rate‐determining step by 0.82 eV. Through integrated in situ Raman spectroscopy and density functional theory calculations, the dynamic U(VI)/U(V) transition is directly captured at Ti‐active edge sites, representing the first mechanistic elucidation of MXene‐based electrocatalysis for uranium. Consequently, the TiMX/CNT‐COOH cathode achieves an impressive uranium extraction capacity of 1,1719.96 mg g −1 with excellent cycling stability. This work offers fundamental insights into the electrocatalytic mechanism and provides a strategic framework for designing large‐scale electroactive materials for uranium recovery from wastewater.

Here, we present a breakthrough in fabricating 3D porous MXene/VMT (3D-TV) electrodes via a facile ice-templating route. This strategy not only effectively mitigates nanosheet restacking but also eliminates the need for costly additives and complex processing. The resulting electrode shows an impressive uranium adsorption capacity of 1852 mg g-1 at -5 V, demonstrating a novel and scalable pathway for electrochemical uranium recovery.

The mining of uranium inevitably leads to the environmental diffusion of wastewater, posing potential threats to biological health. Currently, MXene-based materials exhibit remarkable potential in the field of electro-driven uranium remediation due to their distinctive characteristics. However, their development is hindered by challenges such as susceptibility to oxidation, inadequate stability, and interlayer stacking, which significantly impede challenges to further developments and practical applications. Herein, we report the carbonizion of polyphosphazene derived materials with MXene to fabricate a self-supporting electrode (designated as MPC) for the electrochemical removal of uranium. The sintered carbides act as an armor that supports the pore structure of Ti3C2Tx, thereby enhancing both the stability and oxidation resistance of MPC. Notably, the remarkable specific surface area and "carbon armor" structure contribute to an unprecedented uranium removal capacity of up to 8901.46 mg/g. The relatively low reduction potential (-0.39 V vs. SHE) inhibits the disproportionation reaction of U(V), leading to its reoxidation and ultimately facilitating the formation of a neutral insoluble substance Na2O (UO3 & sdot;H2O)x. This innovative structure is anticipated to challenge traditional synthesis concepts associated with MXene-based electrodes and provide a new paradigm for developing electrochemical uranium removal electrodes.

The two-step vitrification process involves converting high-level waste from a liquid to a solid state through calcination, followed by melting it with the glass matrix material. The calcination process is a critical stage in the vitrification process. This study focused on investigating the thermal decomposition performance and chemical phase changes of power reactor simulated high-level waste during the calcination process. Differential thermal analysis (DTA) was used to assess the thermal decomposition behavior, while scanning electron microscopy (SEM) was utilized to observe the morphology of waste. X-ray diffraction (XRD), Raman spectroscopy, and infrared spectroscopy (IR) were used to determine the phase and structure of the waste during calcination. The results showed that as the simulated waste underwent calcination between 100 and 900 °C, it displayed a darker color and reduced particle size. Below 600 °C, the waste experienced thermal decomposition of nitrate and formate compounds, leading to an approximate 40 wt% weight reduction. At 600 °C, the waste's weight remained constant, and crystalline phases such as ZrxM1-xO2 (M = Ce, Pr, Nd), CeO2, and GdxCe1-xO2 were observed.

Uranium serves as a fundamental element for sustainable nuclear reactor energy. Hence, the extraction and enrichment of uranium from radioactive wastewater is of considerable significance for both nuclear power utilization and environmental protection. In this study, carbon microspheres (TAC) doped with nitrogen, phosphorus and sulfur were derived from ternary polyphosphazene. Subsequently, molybdenum disulfide (MoS2) was integrated using hydrothermal method to fabricate the TAC/MoS2 composite, which is designed for the efficient enrichment and separation of uranyl ions. Owing to the unique two-dimensional structure of MoS2 and its strong affinity towards uranyl ions, the adsorption capacity of the composite was significantly enhanced to 399.6 mg/g within 80 min at pH 5 and 298.15 K. Additionally, the composite exhibited commendable thermal stability and hydrophilicity, declaring that it is an outstanding candidate for the efficient enrichment and separation of uranyl ions. Further mechanism studies through XPS analysis and DFT calculations suggested that the vacancies introduced by MoS2 can generate electrostatic interactions and coordination effects with uranyl ions, thereby improving the adsorption performance of the composite. Due to its straightforward preparation process and exceptional adsorption performance, TAC/MoS2 possesses broad application prospects in the extraction of uranium from wastewater.