BackgroundThe distribution coefficient (Kd) of radionuclides on bentonite is one of the key parameters in the safety assessment of geological repositories for high-level radioactive waste.PurposeThis study aims to reliably predict the Kd values of various radioactive cations in bentonite based on machine learning (ML) models.MethodsBased on the Japan Atomic Energy Agency's adsorption database (JAEA-SDB), 1 240 sets of Kd data of 10 nuclides (Am, Bi, Cm, Cs, Eu, Ni, Pb, Po, Ra, U) in bentonite were collected, and 9 input factors were selected to construct six ML models including random forest (RF) and support vector regression (SVR). The robustness of the ML models was evaluated by Monte Carlo cross-validation (MCCV).ResultsThe validation results indicate that the RF model is the most effective in predicting Kd, achieving a determination coefficient (R2) of 0.902 9 on the training set and 0.728 6 on the test set. Moreover, the RF model accurately reproduce the probability density distribution characteristics of Kd, and pH, initial concentration, ionic strength, and temperature are the main factors affecting Kd.ConclusionResults of this study demonstrate that ML methods, especially the RF model, can rapidly and reliably predict the Kd of multiple radionuclides on bentonite under complex conditions, offering a promising new approach for the safety assessment of radioactive waste disposal repositories.

BackgroundThe detection of trace uranium in water is essential for mitigating health risks associated with uranium exposure. Electrochemical detection presents a promising and efficient approach for the rapid, real-time monitoring of trace uranyl ions (UO?2?) in aqueous environments.PurposeThis study aims to utilize the high specific surface area and the hybridization ability of the surface orbitals of antimonene (AM), a two-dimensional material, to load oligonucleotides on the surface of AM by self-assembly method, and to be used as a specific uranyl ion probe for the electrochemical detection of trace uranium in water.MethodsFirst of all, test samples consisted of Uranyl ion (UO22+), Oligonucleotide, AM, etc., were prepared according to strict processing steps. Then, atomic force microscopy (AFM) and ultraviolet absorption spectroscopy (UVAS) were employed to observe and confirm the successful loading of oligonucleotides on the antimonene surface. Cyclic voltammetry (CV) and electrochemical impedance spectroscopy (EIS) scans were applied to testing that AM-loaded oligonucleotides were less internally resistive and more electrochemically active than individual oligonucleotides or AM. The effects of electrode modification density, temperature, pH, and enrichment time on the detection performance of uranyl ions were further explored using differential pulse voltammetry (DPV) tests. Finally, the sensitivity, specificity, and stability of the electrochemical sensors based on AM-loaded oligonucleotide probes were evaluated.ResultsTest results show that the detection conditions are optimized using the DPV method, and the optimal conditions are found to be a modifier concentration of 0.8 mg?mL-1, an electrolyte pH of 3.0, a temperature of 30 ℃, and an incubation time of 12 min. Under the optimized conditions, the linear range is 1.48×10-8~1.07×10-7 mol?L-1, and the detection limit is 2.99×10-10 mol?L-1, along with good reproducibility, selectivity, and reliability for real-water detection.ConclusionsA novel strategy for constructing high-performance electrochemical sensors for uranyl ion detection using oligonucleotides proposed in this study presents an innovative probe material design scheme to enable the development of simple, efficient, and flexible electrochemical uranyl sensors for practical applications.

BackgroundThe surrounding rock of a repository is the last barrier to block the migration of radionuclides, and there is a potential risk of co-migration of mineral colloid and nuclides in the fissure.PurposeThis study aims to investigate the main components of the surrounding rock mineral (SRM) colloid and its adsorption mechanisms for radionuclide Cs+, as well as to assess the co-migration risk of mineral colloid and Cs+.MethodsFirstly, SRM colloid was prepared from rock samples collected from a cavern disposal repository, and its chemical composition and microscopic structure were characterized using X-ray fluorescence spectroscopy (XRF), X-ray diffraction (XRD), Fourier transform infrared spectroscopy (FT-IR), and scanning electron microscopy (SEM) techniques. Then, batch adsorption experiments were conducted to investigate the effects of contact time, temperature, initial Cs+ concentration, pH value, and coexisting ions on Cs+ adsorption by SRM colloid. Finally, adsorption kinetics, thermodynamics, and isotherm models were used to analyze the adsorption mechanism.ResultsThe characterization results show that SRM colloids are mainly composed of silicate minerals including biotite, K-feldspar, and albite. The adsorption experiments reveal that the saturated adsorption capacity of SRM colloid for Cs+ is 47.13 mg·g-1, and the adsorption process follows pseudo-second-order kinetics (R2=0.999). Under acidic conditions (pH<6) and in the presence of high concentrations of K+ and Ca2+, colloid stability decreases and Cs+ adsorption is significantly inhibited. Conversely, in alkaline environments (pH>6), colloid stability improves and Cs+ adsorption increases to a maximum of 50.7 mg·g-1 at pH value of 8. Interestingly, Mg2+ shows a promoting effect on Cs+ adsorption as its concentration increases, with adsorption capacity reaching 53.1 mg·g-1 at 1 mmol·L-1 Mg2+.ConclusionsThe adsorption of Cs+ by SRM colloid is a monolayer, irreversible chemical process that fits well with the Langmuir isotherm model (R2=0.942). Mg2+ promotes Cs+ adsorption by altering the microstructure of SRM colloid and increasing the interlayer spacing. Experimental results demonstrate that SRM colloid can facilitate Cs+ co-migration, but this risk can be mitigated by modifying the pH of backfill materials to acidic conditions or by using calcium-based materials instead of magnesium-based materials during repository construction.

BackgroundPlastic scintillation microspheres (PSm) are a novel luminescent material with significant potential in the measurement and continuous monitoring of radionuclide activity.PurposeThis study aims to optimize the application of PSm in radionuclide activity analysis by exploring the impact of different measurement conditions on the background count rate and counting efficiency of PSm.MethodsFirstly, PSm samples were prepared by suspension polymerization method, with spherical particles mainly distributed between 95~200 μm in size, and a Hidex 300SL liquid scintillation counter was employed to measure the chemiluminescence and photoluminescence of PSm. Then, the effect of PSm dosage and the type of liquid scintillation vial on the background count rate of PSm were investigated, and the counting efficiency of PSm under different conditions was determined using 14C as a reference radionuclide. Finally, a 20 mL plastic liquid scintillation vial with the lowest detection limit was selected for the activity analysis of 14C solution.ResultsMesaurement results show that weak interference signals from chemiluminescence and photoluminescence are observed when PSm is mixed with water. This interference can be mitigated by keeping the samples in a light-protected environment. The background count rate is positively correlated with the volume of PSm, with larger volumes resulting in higher background counts. The counting efficiency is positively correlated with the height of PSm, with higher heights corresponding to higher counting efficiency. Differences between plastic and glass liquid scintillation vials of comparable specifications are minimal. The minimum detectable activity (MDA) of the 20 mL vial is lower. Analysis of 14C solutions in the range of 574.5 Bq·L-1 to 3 825.6 Bq·L-1 shows good agreement between measured and expected values.ConclusionsThe study demonstrates that the background count rate and counting efficiency of PSm are influenced by various factors such as sample volume. Under fixed conditions, PSm measurements exhibit good accuracy. PSm shows promising potential in radionuclide analysis.

BackgroundLiCl-KCl eutectic is often used as the electrolyte in pyroprocessing of spent nuclear fuel. The corrosion of structural materials by LiCl-KCl eutectic is primarily affected by impurities. Purifying molten salt to reduce its corrosivity is one of the main ways to solve the material corrosion issue in pyroprocessing.PurposeThis study aims to investigate the effect of electrolytic purification on the corrosion of Inconel 600 by LiCl-KCl Eutectic molten salt.MethodsThe purification of LiCl-KCl salt was conducted by electrolysis method in this study and the cyclic voltammetry (CV) test was performed for monitoring the concentrations of impurities during the purification process. Inductively coupled plasma mass spectrometry (ICP-MS) measurement was performed for the salt to examine the impurity elements before and after purification. Inconel 600 alloy samples were subjected to a 500-h immersion corrosion test with purified and untreated LiCl-KCl salts under Ar atmosphere at 773 K. Scanning electron microscopy (SEM) and X-ray diffractometer (XRD) were employed to characterize the morphology and elemental distribution of the post-corrosion samples.ResultsExperimental results show that the metal ion impurities are removed by electrolytic purification with the removal rates of Fe and Cr reach 46.1% and 51.4%. In argon protected environment, samples are corroded by untreated LiCl-KCl salts and LiCrO2 generates on the surface of samples. Samples exposed to purified salts are less corroded than which exposed to untreated salts.ConclusionsElectrolytic purification has a significant removal effect on metal ion impurities in LiCl-KCl salt, and effectively mitigates the corrosion of Inconel 600 by high-temperature molten salt.

BackgroundBoron Neutron Capture Therapy (BNCT) is an emerging radiotherapy technique. However, respiratory motion has a critical impact on the dose accuracy in BNCT treatment of lung cancer.PurposeThis study aims to quantify the dosimetric impact caused by respiratory motion during BNCT treatment of lung cancer.MethodsThis study adopted the Monte Carlo simulation method was adopted to develop a dynamic model that captured the spatiotemporal variations of tumors and organs caused by respiratory motion during lung cancer treatment, and performed dose calculations for BNCT. Firstly, the Multi-function and Generalized Intelligent Code-bench based on Monte Carlo method (MagicMC) was employed to model the adult male phantom provided by Oak Ridge National Laboratory (ORNL). Then, a dynamic dose calculation model was established by incorporating high-order cosine functions that described respiratory motion. Finally, MagicMC was applied to the calculation of the dose errors in tumors and organs resulting from respiratory motion in different directions within three-dimensional space.ResultsThe results indicate that during a respiratory cycle, the tumor in all three motion directions exhibits the largest percentage dose difference at the 50% phase. In the left-right direction (LR), it is 0.310%; in the anterior-posterior direction (AP), it is 5.830%; and in the superior-inferior direction (SI), it is -2.852%. The closer healthy tissues are to the irradiation field, the higher the dose rate they receive. The maximum percentage dose difference for the heart in the LR direction is 2.070%, and the maximum percentage dose differences for the right lung in the AP and SI directions are 4.128% and -11.962%, respectively. During BNCT treatment irradiation, organ motion in the AP direction has the greatest impact on tumor dose, resulting in a dose error of 1.644%. For healthy tissues, the dose errors induced by motion in all three directions remain within ±4%.ConclusionsThe study demonstrates that organ respiratory motion during BNCT treatment for lung cancer affects the doses received by tumors and healthy tissues, the calculation results can provide a reference for precise dose calculation and clinical irradiation dose correction in BNCT treatment of lung cancer.

Background223Ra SPECT/CT (Single Photon Emission Computed Tomography/Computed Tomography) imaging provides critical information for evaluating the efficacy of bone-targeted radiotherapy in patients with metastatic castration-resistant prostate cancer (mCRPC).PurposeThis study aims to explore the optimal imaging parameters for 223Ra SPECT/CT using three standard phantoms: the Carlson model, PET/CT model, and Hoffman 3D brain model.MethodsFirstly, quality control tests were performed on all three phantoms using a GE NM/CT 670 ES system with 99Tcm to establish standard reference images. Then, 223Ra SPECT/CT imaging was conducted by injecting 3.7 MBq of 223RaCl2 into each phantom, with acquisition parameters set at 84 keV and 140 keV energy peaks with 10% windows, using both LEHP and HEGP collimators, and acquisition times ranging from 30 min to 100 min. Finally, image reconstruction was performed using Butterworth filters (fc=0.48, p=10) and three-dimensional ordered subset expectation maximization (3D OSEM) algorithms with 5 subsets and 10 iterations, followed by qualitative assessment of image quality by two experienced nuclear medicine physicians.ResultsThe 223Ra SPECT/CT images acquired with HEGP collimators demonstrate significantly better resolution and contrast compared to those with LEHP collimators. In the Carlson model, the minimum detectable hot and cold zone diameters are both 22.3 mm, with acceptable linearity but suboptimal uniformity. PET/CT model 223Ra SPECT/CT images are acceptable, but there are certain differences between the transverse, coronal, and sagittal image fusion matching degree and attenuation correction dispersion index compared to the 99Tcm SPECT/CT quality control images. The Hoffman brain phantom 223Ra SPECT/CT images show less distinct brain sulci, gyri, and basal ganglia nuclear groups, indicating a need to further improve image quality.Conclusions223Ra SPECT/CT imaging is feasible for clinical applications, with optimal parameters including an 84 keV±20% energy window and HEGP collimation. While the current protocol achieves diagnostic-quality images with a spatial resolution of 22.3 mm for both hot and cold lesions, further parameter optimization is needed to improve image uniformity and detailed visualization of complex structures. These findings provide a foundation for clinical 223Ra SPECT/CT imaging protocols that can enable more accurate assessment of 223Ra therapy distribution and response.

BackgroundBone fracture is an important factor affecting the life quality and mortality of elderly individuals, and its pathogenesis involves the imbalance of bone metabolism maintained by osteoblasts (OB) and osteoclasts (OC). 99Tc-MDP is a drug for the targeted treatment of osteoporosis. While it can directly inhibit OC activity, there have been no in vivo data on its ability to induce OB activity.PurposeThis study aims to evaluate the effect of 99Tc-MDP on osteogenesis in the treatment of osteoporosis.MethodsIn this study, a rat model of osteoporosis after ovariectomy were constructed to reflect the early process of osteoporosis formation, including calcium loss and bone mineral density decrease. Then, experimental rats were randomly divided into 4 groups, i.e., negative control group, model control group, 99Tc-MDP treatment group and zoledronic acid treatment group, with 5 rats in each group (a total of 20 rats). Subsequently, the dynamic changes of osteoblast indexes such as blood calcium level and blood phosphorus content, after 99Tc-MDP treatment were detected at the cellular, metabolic, and genetic levels to evaluate the effect of 99Tc-MDP on osteogenesis in the treatment of osteoporosis.ResultsExperimental comparison results demonstrate that 99Tc-MDP effectively inhibits the osteoporosis process and reverses bone mineral density loss by inducing OB activity, achieving the suppression of bone decline at 4 weeks and returning to the preoperative level at 8 weeks. Although the OB activity induced by 99Tc-MDP is altered to similar levels in OB cells of normal rats, but there is no significant change in the expression of major bone-related genes. The multifactor analysis results suggest that IL-6 can be the key factor and monitoring index.Conclusions99Tc-MDP can stimulate OB activity as a powerful supplement to inhibiting OC activity, which is beneficial to the maintenance of OB/OC homeostasis.

Background7Be and 210Pb are the main radionuclides for monitoring the quality of the radiation environment of atmospheric aerosols, and the internal irradiation caused by their adsorption into aerosols and entry into the human body will be harmful to the human body, so it is of great significance to explore the characteristics and mechanisms of their spatial and temporal distribution and dose contribution.PurposeThis study aims to investigate the differences in the spatial and temporal distributions of atmospheric 7Be and 210Pb and the mechanisms of these differences.MethodsIn this paper, 75 aerosol samples were collected from Nanning city, from January 2015 to December 2017 and from January 2019 to May 2022 using an ultra-high-flow air aerosol sampler, and the activity concentrations of 7Be and 210Pb in the aerosol were measured and analyzed using a high-purity germanium (HPGe) γ-spectrometer. The activity concentration data of 7Be and 210Pb were systematically collected in 17 areas of our country, as well as PM2.5, PM10 and O3 air concentration data from January 2014 to November 2023 in Nanning city. Finally, the Lagrangian inverse trajectory analysis technique based on the Hybrid Single-Particle Lagrangian Integrated Trajectory (HYSPLIT) model, combined with the clustering analysis method based on MeteoInfo software, was employed to analyze the seasonal characteristic control mechanism of 7Be and 210Pb.ResultsThe activity concentrations of 7Be in 41 aerosol samples from January 2019 to May 2022 ranges from 0.27 mBq·m-3 to 11.95 mBq·m-3, with a mean value of (3.68±0.51) mBq·m-3; the activity concentrations of 210Pb in 60 aerosol samples from January 2015 to December 2017 and from January 2019 to February 2021 ranges from 0.23 mBq·m-3 to 4.33 mBq·m-3, with a mean value of (1.51±0.13) mBq·m-3. Combining the 7Be and 210Pb activity concentration data in 17 areas of China, the seasonal distribution of 7Be and 210Pb activity concentrations in the atmosphere of China is characterized by an overall high in winter and spring, and a low in summer and autumn, but 7Be reaches its maximum value in spring while 210Pb in winter. The correlation analysis shows that 7Be has a significant positive correlation with PM2.5, PM10, O3, and has the best correlation with PM10. 210Pb has a significant positive correlation with PM2.5 and PM10 and has the best correlation with PM2.5, while it is not correlated with O3. The traceability analysis of air masses in Nanning city of the 2021 shows that the summer air masses mainly come from the oceanic air masses in the south with lower particle concentration, and the winter air masses mainly come from the continental air masses in the north with higher particle concentration. Combining the 7Be and 210Pb activity concentration data in 17 areas of China, there are effects in the spatial distribution of 7Be and 210Pb activity concentrations. The latitudinal effects: the mean 210Pb activity concentration in the area north of about 40°N (3.20±0.24) mBq·m-3 is about 2.2 times higher than that in the area south of about 40°N (1.48±0.06) mBq·m-3; the mean value of 7Be activity concentration in the area north of about 35°N (7.44±0.26) mBq·m-3 is about 1.9 times of that in the area south of about 35°N (3.92±0.19) mBq·m-3. The altitudinal effect: the 7Be activity concentration is greater at high altitude than at low altitude. Based on the average of winter-spring and summer-autumn activity concentrations of 7Be and 210Pb in 17 regions of the country, it can be estimated that the radiation dose due to 7Be and 210Pb in winter-spring (42.08 μSv·a-1) is about 1.4 times higher than that in summer-autumn (31.00 μSv·a-1). Integration of monitoring data on activity concentrations of 7Be and 210Pb in 17 regions of the country reveals that the annual effective dose due to 7Be and 210Pb in the area north of 40°N (65.48 μSv·a-1) is about 2.0 times higher than that in the area south of 35°N (32.75 μSv·a-1). The mean annual pending effective doses due to 7Be and 210Pb are 2.47×10-3 μSv·a-1 and 33.57 μSv·a-1, respectively.Conclusions7Be and 210Pb are adsorbed on the particulate matter and transported in the atmosphere, and the PM2.5 and PM10 concentrations are low in summer and autumn while high in winter and spring due to the monsoon regulation, so the 7Be and 210Pb show the seasonal distribution characteristics of low in summer and autumn and high in winter and spring. The highest 7Be activity concentration in spring is related to the "spring leakage", while the highest 210Pb activity concentration in winter is mainly influenced by the Eurasian land-based air masses in the north. The latitudinal effect of 210Pb is probably related to the large amount of particulate matter brought by the northwest land-source winds in winter and the combustion process in winter heating system. The latitudinal effect of 7Be may be attributed to the "spring leakage" of 7Be that occurs in mid-latitudes in the spring. The altitudinal effect of 7Be is controlled by the top-down transport of 7Be from the stratosphere to the troposphere.

BackgroundThe presence of technetium in spent nuclear fuel reprocessing waste solution significantly increases the pressure of glass solidification and geological disposal of radioactive waste, and poses long-term potential radioactive hazards to the ecosystem. Therefore, it is necessary to separate and extract it in advance.PurposeThis study aims to propose a process for the extraction and purification of technetium that can be well adapted to the plutonium uranium recovery by extraction (PUREX) process, providing support for engineering applications.MethodsBased on the experimental data pertaining to the extraction and separation of technetium utilizing NTAamide(C8), a refined technetium separation process route containing technetium extraction and stripping with ammonium carbonate was meticulously designed whilst impurity ions were washed by oxalic acid mixed with nitric acid. Subsequently, validation of optimized process route was finalized through a series of cascade experiments with several stages' extraction and washing.ResultsExperimental results show that good washing effect on each impurity ion is achieved when the concentrations of oxalic acid and nitric acid are 0.10 mol·L-1. An exceptional technetium recovery rate of 99.9% is attained after 8-stage extraction and 8-stage washing process while the purification factors of strontium, cesium, zirconium, and ruthenium reach 6.9×103, 7.9×104, 4.3×102 and 45, separately.ConclusionThis optimized process proposed in this study provides technical support for the separation and extraction of technetium in post-treatment plants.

BackgroundHigh dose rate (HDR) brachytherapy is a widely utilized treatment modality in modern clinical brachytherapy. In clinical practice, accurate dosimetric parameters for 192Ir HDR brachytherapy sources are essential. Due to the variation in source designs, each model requires specific dosimetric parameters. Although there is extensive international research on the dosimetric parameters of 192Ir sources, simulation studies focusing on the dosimetric parameters of HDR 192Ir brachytherapy sources produced by the Atomic Hi-Tech (HTA) Co., Ltd are relatively scarce in China.PurposeThis study aims to calculate the dosimetric parameters of a domestically produced HDR 192Ir source by constructing a detailed structural model of the source using Monte Carlo simulation software, based on the dosimetric calculation methods recommended in the TG-43U1 report by the American Association of Physicists in Medicine (AAPM).MethodsBased on the dosimetric calculation methods recommended in the TG-43U1 report by the American Association of Physicists in Medicine (AAPM), Monte Carlo software was used to simulate the 192Ir brachytherapy source, and a detailed structural model of the domestically produced high dose rate 192Ir brachytherapy source was established within the Monte Carlo software to accurately conduct the simulation. Then, the dosimetric parameters calculated in the simulation such as the dose rate constant, air kerma strength per unit activity, radial dose function, and anisotropy function were calculated in the simulation and compared with that reported in the literature.ResultsSimulation results show that the simulated dose rate constant is found to be 1.105 cGy·h-1·U-1, with a difference of less than 1.2% compared to values reported in the literature. Additionally, the air kerma strength per unit activity is calculated to be 9.788×10-8 U·Bq-1, with a discrepancy of 0.23% compared to the literature values. Furthermore, the dosimetric parameters for the radial dose function and anisotropy function obtained in this study show a high degree of consistency with corresponding data from existing literature.ConclusionsThe domestically produced 192Ir source model established using the Monte Carlo software demonstrates good consistent dosimetric parameters with the literature-reported dosimetric parameters, indicating that this model can be used for clinical practice applications of domestically produced 192Ir sources and has certain guiding significance.

Background14C has become the largest contributor of annual effective dose to the surrounding public in radioactive effluent during normal operation of nuclear power plants whilst 14CH4 is the largest and most chemically stable of the 14C hydrocarbons. However, an effective treatment method for 14C in the form of hydrocarbons has not yet been established.PurposeThis study aims to investigate the treatment performance of a non-thermal plasma/catalytic coupling system for low concentrations of 14CH4.MethodsConsidering the optimization of radiation protection, 12CH4, which possesses the same physicochemical properties as 14CH4, was selected as the experimental subject. The Pd/γ-Al2O3 catalyst were prepared by segmental heat treatment and wet impregnation. The discharge behaviors of the plasma were analyzed by the Lissajous figure. The CH4 treatment performance of the non-thermal plasma before and after the introduction of the catalyst under different discharge voltages, gas flow rates and temperatures were analyzed by the constructed experimental system. The microstructural changes of the catalyst were analyzed by N2 adsorption-desorption isotherm, X-ray diffraction (XRD) and X-ray photoelectron spectroscopy (XPS).ResultsThe introduction of Pd/Al2O3 catalyst can significantly improve CH4 treatment performance in non-thermal plasma, and the CH4 treatment efficiency and CO2 selectivity can reach 100% and 83.7% by increasing the discharge voltage, lowering the reaction temperature and reasonably adjusting the gas flow.ConclusionThe non-thermal plasma and the Pd/Al?O? catalyst exhibit a synergistic effect. The Pd/Al?O? catalyst can reduce the reaction barriers, optimise the reaction paths and significantly improve the CH4 treatment performance of the plasma. Furthermore, the plasma-excited reactive species are conducive to the formation of PdO, which is the key reactive phase in the catalyst.

BackgroundIn uranium mining and smelting, a significant amount of uranium-containing wastewater is generated, which can easily cause groundwater pollution. In polluted groundwater, uranium often combines with carbonates to form more diffusive and migratory uranyl carbonate complexes, making uranium removal more difficult.PurposeThis study aims to consider the performance of a three-dimensional electrochemical system of rGO/BB hydrogel particle electrodes on uranyl carbonate, and implement an efficient method for removing uranium-containing solutions.MethodsFirstly, graphene oxide (go) samples were prepared by modified Hummers method, then, the effects of electrolyte (sodium nitrate) concentration, applied voltage, electrode spacing, pH value, particle electrode dosage, and initial uranium concentration on the adsorption performance of uranyl carbonate ions by a three-dimensional electrochemical system were investigated, and the recycling performance of the system was studied. Finally, the adsorption mechanism of uranium was analyzed by scanning electron microscopy (SEM), Fourier transform infrared spectroscopy (FT-IR), Raman spectroscopy, and X-ray photoelectron spectroscopy (XPS).ResultsExperimental results show that under the conditions of pH value of 4~8, electrode spacing of 4 cm, voltage of 5 V, and rGO/BB hydrogel dosage of 90 mg, 200 mL of uranyl carbonate with a concentration of 1~1 000 mg?L-1 exhibits good adsorption efficiency. Even when the uranyl carbonate concentration is 1 000 mg?L-1, the adsorption rate can reach 87.56% in 14 h. After five cycles of adsorption and desorption, the adsorption rate remains above 87%, thus showing good recycling performance. The addition and use of rGO/BB hydrogel particle electrodes significantly improves the adsorption capacity of the electrochemical system for uranium, in which carboxyl and hydroxyl groups play a major role in uranium adsorption.ConclusionsThe results of this study show that the three-dimensional electrochemical system based on rGO/BB hydrogel particle electrodes has great application potential for removing uranyl carbonate ions.

BackgroundReprocessing technology is recognized internationally as one of the most promising technologies to realize the closed cycle of nuclear fuel. Shanghai Institute of Applied Physics (SINAP) of Chinese Academy of Sciences (CAS) has been focusing on the development of this technology, as well as the corresponding support systems engineering in the past decade. The hot cell is an important guarantee for the practical application of nuclear fuel reprocessing technology, therefore, a remote operation system suitable for reprocessing post-treatment process equipment is developed so that the pyroprocessing experiments of molten salt can be conducted in hot cell.PurposeThis study aims to evaluate equipment remote operation and pyroprocessing verification in hot cell.MethodsMain focus of this study was the fluoride volatility and vacuum distillation process of molten salt reactor fuel salts, and the remote operation evaluation of hot cell equipments and pyroprocessing verification experiments were carried out using multi-view coordination. The workload of the operation of the fluoride volatility and vacuum distillation process units was analyzed and evaluated, and the operation efficiency was obtained based on the frequency and time spent on the basic movements of the manipulator in pyroprocessing verification experiments. On this basis, experiments were conducted on the uranium fluoride volatility in the LiF-BeF2-ZrF4-UF4 salt, and vacuum distillation process in LiCl-KCl, LiF-NaF-KF molten salts. Then, before and after the fluoride volatility experiments, inductively coupled plasma mass spectrometry (ICP-MS) was used to determine the uranium content in molten salt, and the conversion rate and reaction rate of uranium were calculated. The uranium content in the base solution downstream of adsorption column was analyzed to obtain the recovery rate of uranium product. Finally, the molten salts evaporation was calculated by the residual mass after the vacuum distillation experiments. The evaporated salts were collected through the condensing cover, and the collection rate was calculated.ResultsThe results of reprocessing validation experiments in hot cell show that the workload of feed and discharge in the operation unit is large than 2.0 on the basis of reasonable arrangement of process equipments in hot cell. The load values for the on/off operation are relatively small, with values of 0.07 and 0.14, respectively. In the operations of processes, due to the various types and frequency of the manipulator operation, the feed and discharge units take a long time with a time-consuming of 20 min and 19 min respectively. The operational efficiency of the pyroprocessing in the hot cell can be improved by optimizing processes and reducing unnecessary operations. A uranium conversion rate of 99.8% and recovery rate of over 99% in molten salt are achieved in he uranium fluoride separation experiment. By improving distillation temperature and sealing of components, a higher evaporation rate and a 100% recovery rate of molten salt distillation are achieved in the vacuum distillation experiments.ConclusionsThe designed small-scale fluoride volatility and vacuum distillation devices can be used for remote operation and process experimental research in the hot cell, and the experimental results meet the key process targets. The research work can provide a reference basis for the pyroprocessing of real spent fuel in hot cell.

BackgroundRubber-based nanocomposites have become a research focus in the nuclear industry due to their wide application in wearable radiation protection devices.PurposeThis study aims to explore the γ-ray shielding mechanism of composite material of Bi2WO6 nanoparticles and natural rubber (NR), so as to provide theoretical support for the further materials preparation of low toxicity, light weight and high efficiency radiation shielding.MethodsBismuth tungstate (Bi2WO6) nanoparticles synthesized via hydrothermal process, additionally, WO3 and Bi2O3 particles were prepared by ball milling method. Then, these particles were filled into natural rubber (NR) at the mass fraction of 30% to fabricate three composites: NR/Bi2WO6, NR/WO3 and NR/Bi2O3. Finally, laser particle size analyzer, X-ray diffraction analysis (XRD) and field emission scanning electron microscope (FE-SEM), etc., were employed to access the mechanical properties of NR/Bi2WO6, and the γ-ray shielding effect was evaluated on the basis of the γ-ray shielding effects of Bi2WO6, WO3 and Bi2O3.ResultsThe results show that the NR/Bi2WO6 nanocomposites achieves a γ-ray shielding factor of 13.6% for 59.5 keV (241Am point source), which is significantly higher than NR/WO3 (7.4%) and NR/Bi2O3 (9.2%). Furthermore, a comparison of WO3 and Bi2O3 indicates that the interlayer effect of the Bi2O22+ and WO42- layers in the Bi2WO6 cell is conducive to increasing the probability of collisions between γ photons and extranuclear electrons.ConclusionsThe γ-ray shielding performance of NR/Bi2WO6 composites is significantly improved by the boundary complementary effect of both K and L layer absorption edges of W and Bi elements existed in Bi2WO6 nanoparticles, which enhances the attenuation efficiency of NR/Bi2WO6 to the γ-ray.

BackgroundThe C50 concrete is used as a concrete structural materials for engineering disposal units in cavern-type low and intermediate radioactive waste repositories in order to meet the needs of nuclear power plant operation and decommissioning.PurposeThis study aims to prepare C50 concrete material and investigate its adsorption performance for radioactive nuclides Ni2+ and Cs+.MethodsFirstly, according to the design requirements of C50 concrete, C50 concrete specimens were prepared, and then crushed and ground into powder with a particle size less than 75 μm as experimental samples. Then, series of characterization analysis were performed to observe the crystal structure and content of the C50 concrete sample to reveal the main components of the C50 concrete that comprised silicate minerals with high aluminum content and carbonate minerals with high calcium content. Finally, adsorption experiments were carried out to examine the adsorption performance of C50 concrete for radioactive nuclides Ni2+ and Cs+. Considering the widespread presence of ions in groundwater, the effect of the coexistence of different ions on the adsorption of Cs+ and Ni2+ in the C50 concrete was tested, and the effect of temperature on adsorption was investigated with concern of the radiation exotherms of radioactive waste.ResultsThe analysis results show that the main components of the C50 concrete are SiO2, CaO, and Al2O3 with and specific surface area of 2.786 m2·g-1, and irregular polyhedral and lamellar structures with strong amorphous characteristics are noted at the microscale. The adsorption experiments reveal that the C50 concrete have a good adsorption capacity for Cs+ and Ni2+. The Kd of Cs+ adsorption on the C50 concrete reaches 19.400 L·mg-1 with adsorption capacity of 0.316 mg·g-1 whilst the Kd of Ni2+ adsorption on the C50 concrete reaches 465.142 L·mg-1 with adsorption capacity of 96.375 mg·g-1. With the increase of initial metal ion concentration, the Kd value of Cs+ adsorption on the C50 concrete gradually decreases whilst the adsorption capacity increases. The Kd of Ni2+ increases steadily after a gradual decrease, whereas the adsorption capacity gradually increases with the increase of initial metal ion concentration. Experiments results on the influence of environmental factors demonstrate that the effect of pH on the adsorption of Cs+ in the C50 concrete is relatively small, and the order of inhibition of ions on adsorption is K+ > Ca2+ > Mg2+ > SO42-> Cl-> NO3-. Increasing the temperature leads to a slow decrease in the absorptivity. Meanwhile, as for the adsorptivity of Ni2+ in the C50 concrete increases with pH, and the order of inhibition of ions on adsorption is Mg2+ > Ca2+ > SO42- > CO32-. The adsorptivity of the C50 concrete for Ni2+ increases slowly with an increase in the temperature.ConclusionsThe results of this study provides the basic data for the engineering geological disposal of nuclear waste.

BackgroundCement serves as an essential cementitious material required for the construction of repositories for radioactive waste. Over the operational lifespan of such repositories, environmental CO2 infiltrates the cement, leading to its carbonization and alteration of its physicochemical properties. This, in turn, affects its efficacy in blocking radionuclides.PurposeThis study aims to assess the long-term safety implications of cement carbonization on radioactive waste repositories.MethodsThe carbonization patterns of cement and its adsorption capabilities regarding the fission nuclide 137Cs was taken as investigating object. Characterization techniques, such as X-ray fluorescence spectrometer (XRF), X-ray diffractometer (XRD), scanning electron microscope (SEM), Fourier transform infrared spectroscopy (FT-IR), etc., were employed to analyze the changes in physicochemical properties of cement before and after carbonization. Furthermore, batch adsorption experiments were conducted to examine the adsorption behavior of 137Cs in carbonized cement, elucidating the impact of carbonization on cement's adsorption performance.ResultsAnalysis results reveal that the carbonization process in cement primarily involves the conversion of hydration products such as Ca(OH)2 and hydrated calcium silicate into CaCO3, resulting in an increase in the specific surface area of cement with higher degrees of carbonization, hence significantly enhance the adsorption capacity for 137Cs due to carbonization. Interestingly, the adsorption capacity exhibits an initial increase followed by a subsequent decline with increasing degrees of carbonization, surpassing that of non-carbonized cement.ConclusionResults of this study implicate that cement's adsorption of 137Cs operates via chemical single-layer adsorption, and the mechanism remains unchanged by carbonization.

BackgroundAccumulation of α-synuclein is a major hallmark of Parkinson's disease (PD). The development of PET tracers to visualize aggregated α-synuclein is useful for early diagnosis and treatment of PD.PurposeThis study aims to design and synthetize a novel PET tracer, i.e., 18F-YM, for for alpha-synuclein PET imaging, and conduct preliminary biological evaluation.MethodsFirstly, based on benzothiazole scaffolds, 2-((3-fluorobenzyl)thio)-6-(3-[fluorine-18] propoxy)benzo[d]thiazole, a small molecule compound, denoted as 18F-YM, was prepared as PET tracer and labeled using Cu(II) mediated radiofluorination methods. The imaging properties of the tracer were primarily evaluated through PET imaging at A53T mice and normal mice. Additionally, the imaging properties of the tracer were also investigated through biodistribution experiments as well as ex vivo autoradiography and pathological analysis. Then, through chemical synthesis, compounds Sn-YM and 19F-YM were obtained, and the Sn-YM was labeled with 18F using organic tin fluoride method whilst the resulting product 18F-YM was verified by high performance liquid chromatography. The in vitro stability and octanol-water partition coefficient of 18F-YM were determined. Finally, small animal Micro PET imaging was employed to assess the affinity of 18F-YM for α-synuclein, and autoradiography, pathological analysis, and biodistribution were used to validate the results of small animal Micro PET imaging.ResultsObserved results show that 18F labeled small molecule compound is prepared in nearly 1 h with an undecayed yield greater than 10% and a radiochemical purity greater than 95%. In vivo PET imaging of 18F-YM reveals that more radioactivity is significantly detected in the brains of A53T mice than those of normal mice after administration of 18F-YM. Quantitative analysis shows that the uptake values in the brain of A53T mice and normal mice are (2.35±0.06) %ID/g and (1.38±0.15) %ID/g, respectively while ex vivo autoradiography and histological examination confirm the possibility of detection of aggregated α-synuclein in thalamus, substantia nigra and striatum using 18F-YM. Furthermore, a biodistribution study in normal mice reveals that 18F-YM can be quickly cleaned from the brain of normal mice, indicating that the non-specific binding of 18F-YM in the brain is low, which allowed for obtaining good contrast images.ConclusionThis preclinical study demonstrates that the benzothiazole analog, 18F-YM, owns preferable imaging prosperities, hence be a new candidate for α-synuclein PET imaging.