Journal of Radioanalytical and Nuclear Chemistry最新文献

To extract the fission product Gd from spent nuclear fuel, this paper investigates the thermodynamic and kinetic parameters of Gd³⁺ on different electrodes (Mo and Zn). On a Zn membrane electrode, the standard Gibbs free energies of formation ((Delta {G}_{text{f}}^{0})) of four Zn-Gd alloys were calculated. On a liquid Zn electrode, the exchange current density (({j}_{0})) and saturation solubility (({X}_{text{Gd}left(text{Zn}right),text{max}})) were calculated, and the apparent morphologies and elemental compositions of the Zn-Gd alloys were analyzed by SEM–EDS and XRD. Additionally, through ICP-AES analysis, the extraction efficiencies of Gd³⁺ from the system were found to be 99.15 ± 0.39, 99.09 ± 0.21, and 99.58 ± 0.16%, respectively.

The radiochemical methods for the characterization of hard-to-measure radionuclides are often complex, laborious and expensive, especially in the case of solid samples. The present work aimed at developing a new optimized method based on Cherenkov counting for the determination of ³⁶Cl in activated concrete. The primary goal consisted in improving simplicity and cost-effectiveness with respect to currently available methods without sacrificing performance. A chemical recovery of ≈ 90% and high decontamination from interferents were demonstrated. The method also proved to be scalable to large sample masses for clearance purposes, thereby achieving a detection limit < 5 mBq g^–1.

The single- and multi-stage Cs⁺, Sr²⁺and Co²⁺ ions adsorption by a sorbent based on Zr–Ca–Mg phosphates from single-component and multi-component solutions depends on background electrolytes and the V/m ratios were studied. The maximum removal efficiency of 69–95% with a single-stage solution purification was achieved at V/m 125–250 ml/g. Multi-stage purification of solutions in 4 stages reached to increase the removal efficiency up to 100% and significantly reduced adsorbent consumption. The residual metal concentrations after 4 cycles multi-component solution purification were 0.004–0.166 versus 0.71–1.18 mg l⁻¹ after a single purification with twice sorbent dosage. There was a predominant Co²⁺ ions removal during a multi-component solution purification, in the first cycle and Cs⁺ at further stages. Co²⁺ and Sr²⁺ ions adsorption remained at a high level in the presence of H₃BO₃ and Na⁺ ions, while the Cs⁺ ions removal was least affected in presence of H₃BO₃ and Ca²⁺ ions.

With the development of nuclear-powered mobile platforms, the demand for lightweight and high-efficiency neutron shielding materials is becoming increasingly urgent. In this study, the modified-SmB₆ + HDPE composite was prepared. We calculated its shielding performance against neutrons of different energies using MC simulation. The thermal neutron dose shielding rate of this composite with a thickness of 1 cm was increased by 5.8% and 2.4% respectively compared with Sm₂O₃ + HDPE and boron polyethylene of the same thickness. The dose shielding rate of the material with a thickness of 10 cm for ²⁴¹Am-Be neutrons was 86.18%. The experimental data were in good agreement with the simulation data, with a minimum deviation of 7.07%. This composite has broad prospects as a lightweight neutron shielding material.

Feed-forward neural networks with highly pre-processed inputs are proposed as an approach to automatically identify the presence of a wide variety of radioisotopes in complex samples. The process links characterised peaks from gamma spectroscopy to radionuclide emission lines. By inputting relational parameters between spectral peaks and decay data, the neural network selects between two possible solutions with 88% accuracy. The developed neural network demonstrates radionuclide insensitivity and can correctly identify radioisotopes that were not present in its training dataset with comparable accuracy.

Lignin, an abundant biopolymer in lignocellulosic biomass, possesses intrinsic antioxidant properties. In this study, lignin was isolated from black liquor, a by-product of second-generation bioethanol production, via acid precipitation and irradiated with gamma rays at 25, 50, and 75 kGy, using the dose rate of 3 kGy per hour. Structural and functional changes were examined, including free radical content, functional groups, degradation products, thermal stability, and antioxidant activity. Gamma irradiation induced partial depolymerization, increased radical species and oxygenated groups, and reduced thermal stability. The 2,2-diphenyl-1-picrylhydrazyl (DPPH) assays revealed dose-dependent IC₅₀ values, with the highest antioxidant enhancement at 75 kGy, classified as moderate by the antioxidant activity index.

Residual moisture in molten salts hydrolyzes to form hydroxide (OH⁻) ions. Hydroxide dramatically increases the corrosivity of the molten salt and can threaten the integrity of materials in molten salt nuclear reactors (MSRs) and solar-thermal plants. Despite its importance, a reliable, real-time method to quantify OH⁻ in molten salts has yet to be developed. Here, we present a novel method for detecting and quantifying hydroxide ions in molten salts using near-infrared (NIR) spectroscopy. We also demonstrate the use of this in-situ NIR spectroscopy to monitor the removal of moisture from molten salts via chemical and electrochemical moisture removal processes. Additionally, the molar absorption coefficient for the 2ν_OH vibration in molten LiCl–KCl at 773 K was determined and reported based on the Beer–Lambert law. This study not only establishes a powerful spectroscopic tool for OH⁻ detection but also for real-time monitoring and control of salt purity in various molten salt applications.

The study reports enhanced thermoluminescence response of dysprosium-doped aluminium–lithium-zinc-borate glasses for dosimetry applications. The glasses were prepared following the melting quenching process. X-ray diffraction (XRD) analysis demonstrates the absence of sharp Bragg peaks, confirming the amorphous nature of the samples. The density of the glass samples decreases from 5.60 to 1.56 g/cm³, and the molar volume increases from 12 to 49.18 cm³/mol with an increase in dysprosium concentration. Consistent particle distribution and homogenous surface morphology were shown by field emission scanning electron microscopy (FESEM) examination, suggesting a stable glass matrix structure appropriate for dosimetric applications. Fading of 7%, 10%, and 11% after 20 days, 30 days, and 50 days. The effective atomic number of the samples increased from 7.10 to 9.97 eV with a significant increase in sensitivity of 38.1–88.8 (Gy/nC/g) with increased dysprosium. The transitions peak photoluminescence emission shows: 4F9/2 → 6H15/2, 4F9/2 → 6HJ/2, and 4F9/2 → 6H13/2 at concentrations of 1.5 mol%, 2.5 mol%, 3.4 mol%, and 4.8 mol% Dy³⁺ observed. The samples with dysprosium show good reproducibility. Trap depth characteristics, an optimal energy level, and stability, thus reinforcing the glasses’ suitability for radiation dosimetry applications.

In nuclear reactors, nuclear fuel undergoes a complex fission process, Ce and Gd are fission products generated directly from the fissile materials. Generally, Ce and Gd are difficult to separate individually due to their similar chemical properties. In this study, a method to achieve the separation and co-reduction of Ce and Gd in LiCl–KCl molten salt system at 873 K is presented. It is found that Gd could be separated by electrolysis at −2.0 V, and Ce and Gd could be co-reduced at −2.4 V. A relatively pure product could be obtained by the potentiostatic electrolysis (PE) mode.

Traditional polymer-assisted deposition has been shown to produce highly uniform thin films of metal oxides, including actinide oxides. While producing thicker films for nuclear targets is possible through repeated coating application, we exchanged the dissolved metal species with nanoparticles to maximize the thickness that can be achieved with an individual layer. Using CeO₂ nanoparticles in a polyethyleneimine matrix, we produced targets with single-layer areal densities of 0.22 ± 0.01 mg·cm⁻² (1σ) and thicknesses of 690 ± 80 nm (1σ). A custom 3D-printed spin coating chuck attachment with an inlay improved target homogeneity and will streamline future work with radioactive materials.