This paper presents a comprehensive review and case-based application of modeling radiolysis-enhanced corrosion phenomena in the primary circuit of pressurized water reactors(PWRs),with particular emphasis on the theoretical development and engineering deployment of the LwrChem simulation tool recently introduced in China.Radiolysis of reactor coolant,driven by gamma and neutron fluxes in the reactor core,produces chemically active species that influence redox conditions and,in turn,affect the corrosion potential of structural materials.The LwrChem model couples detailed radiolysis kinetics,electrochemical equilibrium calculations,and spatially resolved three-dimensional core physics to simulate coolant chemistry evolution with high accuracy and strong spatial fidelity.In this study,a typical four-loop PWR was modeled using a 178-compartment spatial decomposition to capture coolant behavior across core and non-core regions.Two distinct operational scenarios were investigated:one under nitrogen-free conditions,and the other reflecting nitrogen-pressurized conditions representative of volumetric control tank(VCT)operation with(7-20)×10-6 N2 concentrations.Simulation results demonstrated that,under standard hydrogen injection conditions(2.76×10-6),LwrChem successfully predicted steady-state hydrogen peroxide concentrations near 80×10-9 in high-irradiation zones and electrochemical corrosion potential(ECP)values ranging from-740 mV to-650 mV vs.standard hydrogen electrode(SHE).These predictions were validated against experimental data with deviations within±50 mV,confirming model accuracy.When nitrogen was introduced into the system,the model captured significant hydrogen consumption(e.g.,500×10-9 reduced to 346×10-9)and the formation of ammonia(NH3)concentrations up to 3 500×10-9,indicating a substantial shift in the redox environment.The appearance of oxidizing species such as NO2 suggested a potential NH3/NO2 synergistic mechanism in accelerating corrosion,although NO2 was not explicitly included in the electrochemical model.Moreover,spatial heterogeneity in ECP values across fuel assemblies was revealed,with lower potentials observed at the top of high-power bundles,highlighting the importance of axial redox variation in understanding localized corrosion and crud-induced power shift(CIPS)mechanisms.The study further introduces the integration of LwrChem with the SIMULATE5 core simulation platform,enabling high-resolution,node-level corrosion risk assessment based on actual power and radiation field distributions.This coupling improves predictive capability for identifying oxidation hotspots and supports targeted water chemistry control strategies,including hydrogen injection optimization.Overall,this work represents the first in-depth engineering validation of LwrChem under Chinese and the first systematic modeling of nitrogen pressurization effects on engineering application of coolant chemistry and corrosion.The model demonstrates strong scalability,broad applicability,and high reliability under extreme conditions,including high temperature,high pressure,and mixed-radiation fields.These capabilities position LwrChem as a valuable computational tool for predicting radiolysis-affected corrosion behavior,supporting nuclear power plant life extension,improving coolant chemistry management,and contributing to the safe,efficient development of advanced reactor technologies.
As one of the most notorious long-lived radionuclides with strong radioactivity, Cs+ is particularly difficult to sequester due to its alkali-metal-ion chemical properties. Therefore, selective Cs+ adsorbents are urgently needed in radioactive wastewater treatment. Stabilized potassium hexacyanoferrates(Ⅱ) have proven effective for selective Cs+ removal; however, their practical application is severely constrained because they are typically obtained as nano- or microparticles that are extremely difficult to granulate. When granulated or supported particles are subjected to water flow, fine solids tend to be washed away, significantly increasing the radioactivity of the effluent due to the entrained Cs+-laden particulates. To address this issue, composite materials with high selective Cs+ uptake were prepared by incorporating stabilized potassium hexacyanoferrates(Ⅱ) into porous spherical activated carbon. The spherical activated carbon support was derived from strong-acid ion-exchange resin beads through calcination and activation. Using a rationally designed procedure, co-precipitation reactions that yield stabilized potassium hexacyanoferrates(Ⅱ) were conducted within the internal pores of the spherical carbon. The resulting composites were characterized by XRD, N2 adsorption-desorption(BET method), SEM-EDX, and ICP-OES/MS. The results show that the specific surface areas of the samples exceed 700 m2/g, with micropores, mesopores, and macropores all being present. It can be safely concluded that the stabilized potassium hexacyanoferrates(Ⅱ) were successfully embedded in the inner crevices of the porous spherical activated carbon, and the composites exhibited excellent selective Cs+ capture ability, with distribution coefficients(Kd) of 344170, 57919, and 42540 mL/g in the presence of 0.1 mol/L Na+, K+, or H+, respectively. Under strong acid interference(1.0 mol/L HNO3), the sample using copper-stabilized potassium hexacyanoferrate(Ⅱ) as the active ingredient achieved the highest Kd of 14692 mL/g. In the presence of concentrated K+(1.0 mol/L), the cobalt-stabilized analogue gave a Kd of 92872 mL/g. The outer shell of the porous spherical carbon is permeable to water and ions but impermeable to the stabilized hexacyanoferrate(Ⅱ) nanoparticles. Hence, the active sites remain fully accessible to Cs+ in solution without being washed away, and the treated liquids remained completely colorless. Moreover, the smooth shell endows the composites with good abrasion resistance and crack-proof properties, preventing pulverization or scaling-off during solid-liquid contact. These characteristics make the prepared materials a promising candidate for treating real radioactive wastewater.
During the pyrochemical electrorefining process of spent nuclear fuel,fission product Ce3+continuously accumulate in the LiCl-KCl eutectic molten salt,altering its original physicochemical properties and potentially affecting the overall electrorefining efficiency and system stability.However,systematic research on the physicochemical parameters of the LiCl-KCl-CeCl3 molten salt system remains limited.In this study,the effects of low-concentration CeCl3 addition(x(CeCl3)=5.0%,the same below)on the key engineering properties of LiCl-KCl molten salt were comprehensively investigated within a temperature range of 673-873 K.This work aims to establish an accurate and reliable evaluation framework combining experiments and simulations to provide crucial data and theoretical support for dry reprocessing technologies.Experimentally,three fundamental macroscopic engineering properties were systematically measured under a high-purity argon atmosphere to establish a robust dataset.Specifically,the molten salt viscosity was determined using the rotational method with a high-precision coaxial cylinder viscometer,which effectively captured the fluid resistance.The conductivity(σ)was obtained via the continuous variation conductivity cell constant(CVCC)method employing an advanced LCR bridge to eliminate polarization effects.Furthermore,the initial crystallization temperature,reflecting the solid-liquid phase equilibrium,was accurately recorded through the cooling curve method.To provide deep theoretical insights at the microscopic level,advanced computational simulations were implemented in parallel with the experimental procedures.First-principles molecular dynamics(FPMD)simulations based on density functional theory were performed using the Vienna Ab-initio Simulation Package(VASP)to reveal the quantum-mechanical features.Moreover,a machine learning deep potential molecular dynamics(DPMD)model was trained using the DeePMD-kit package based on the FPMD data,enabling large-scale and long-term simulations to describe the local structures,coordination numbers,and precise self-diffusion coefficients of the multi-component system.The experimental and simulated results provid detailed insights into the macroeconomic variations and microscopic structures.The viscosity of the system decreases nonlinearly as the temperature increases due to accelerated ionic motion,whereas the incorporation of CeCl3 significantly raises the viscosity;at 673 K,the viscosity of the mixture containing 4.0%CeCl3 reaches 8.06 mPa •s,approximately twice that of the pure eutectic salt.The electrical con-ductivity exhibits an upward trend with rising temperature,but decreases slightly with the addition of CeCl3,with the maximum reduction restricted within 0.2 S/cm due to the larger radius and greater migration resistance of Ce3+.Interestingly,both experimental and simulated data manifest a subtle anomalous increase in conductivity at a tiny dosage of 0.5%CeCl3.Furthermore,as the CeCl3 content increases,the initial crystallization temperature decreases initially and then rises,reaching a minimum turning point near 2.0%LiCl-KCl-CeCl3(620 K).Structural analysis derived from DPMD confirms that Ce3+possesses a remarkably strong coordinating capability,with the fractions of 6-and 7-coordinated polyhedral configurations increasing continuously,leading to a locally ordered structure.In conclusion,this study successfully demonstrates that the accumulation of Ce3+leads to increase flow resistance and slightly impeded current efficiency due to its powerful local structural coordination and high diffusion activation energy.The simulated viscosity via the Stokes-Einstein equation shows an excellent agreement with experimental values,with an average prediction error below 10%,validating the reliability of the DPMD potential model.The integrated experimental-computational framework and the validated database established in this work provide essential theoretical guidance and critical engineering parameter optimization for the design and stable operation of high-temperature pyrochemical reprocessing facilities.
In this paper, activated carbon was selected as the adsorption material to study the adsorption and desorption performance of krypton(Kr) and xenon(Xe) gases at various temperatures. A chromatographic quantitative loop tail gas collection column was prepared to achieve the capture of krypton and xenon in the tail gas of the chromatographic quantitative loop, by using activated carbon. The chromatographic quantitative ring tail gas collection column can operate at liquid nitrogen temperature and is used to capture the krypton and xenon gases in the tail gas of the chromatographic quantitative ring, and the lifespan of the gas collection column is no less than 500 injections. Two chromatographic quantitative loop tail gas emission modes, named venting mode and trapped mode were proposed. The emission mode that tail gas directly discharged into the atmosphere is called venting mode, and the emission mode that tail gas passed through the collection column and discharged into the atmosphere is called trapped mode. These two emission modes were selected to study the influence of the tail gas collection column on the performance of the gas chromatograph by comparing the gas flow rate, the retention time, half peak width, peak area response, and simulated samples measurement of noble gas components. There is no significant effect on the carrier gas flow rate under both modes. The absolute value of relative deviation of retention time in these two modes is no more than 0.3%, the absolute value of relative deviation of peak area response in these two modes is no more than 1.0%, and the relative standard deviation(sr) value of retention time and peak area response in these two modes is no more than 0.7%(n=7). Meanwhile, two types of standard curves of various noble gases in venting mode and trapped mode, two different emission modes are prepared, and all the linear correlation coefficients of standard curves of various noble gases are more than 0.999. Under the two tail gas emission modes, two types of standard curves in different emission modes were selected to measure three groups of identical samples. It shows that the absolute values of the measurement relative deviation of the same sample under the two tail gas emission modes are no more than 1.1%. The installation of the tail gas collection column has no significant impact on the quantitative analysis of the chromatograph. One set of standard curves of various noble gases can also be used under both tail gas emission modes, and there is no significant difference in the quantitative analysis results.
In response to the need for minimizing the volume of radioactive waste during the operation and decommissioning of nuclear facilities, membrane distillation technology has attracted much attention due to its high rejection potential for nonvolatile nuclides, but the mechanism of nuclide rejection reduction in membrane distillation process is still lacking. The effects of feed temperature and feed flow rate on the nuclide rejection of PTFE, PVDF and PP hydrophobic membranes in membrane distillation for concentration of radioactive waste liquid were studied systematically by using a self-made air-gap membrane distillation unit. The mechanism of non-volatile nuclides across membrane and the quantitative relationship between rejection and permeate were revealed by microscopic characterization, rejection experiments and contact angle measurements. The results show that all three membranes exhibit high rejection rates, and the average rejection rates of PTFE, PVDF and PP membranes are 99.994%, 99.969% and 99.886%, respectively. When the flow rate increases from 80 L/h to 180 L/h, the rejection rates of the three materials decrease to 99.979%, 99.928% and 99.453%, respectively. The mechanism analysis shows that the non-volatile nuclides are mainly due to the permeation and migration of liquid droplets through membrane pores after membrane pores are wetted. When the flow rate increases, the pressure of liquid on the membrane surface increases, resulting in the increase of permeation flux of PTFE, PVDF and PP membranes. Under the same flow rate change, the permeation fluxes of PTFE, PVDF and PP membranes increase by 13.43, 8.26 and 34.16 times respectively. PP membrane has more tear structure changes due to uneven pore structure, and the change is most significant. In addition, the durability of hydrophobic performance is a key indicator worthy of attention in the engineering application of this technology in the field of radioactive waste concentration. The decrease of membrane hydrophobicity will lead to the decrease of liquid wetting pressure on the membrane surface, resulting in more feed liquid penetrating into the condensate, which will lead to the decrease of rejection rate. Durability test shows that PTFE membrane keeps good hydrophobic property under long-term high temperature and high salt scouring, while PVDF membrane loses hydrophobic property after 25 hours. Based on the above research, membrane distillation has excellent rejection performance for nonvolatile nuclides, but it is significantly affected by the uniformity of membrane pore structure and hydrophobic durability. Nuclides migrate across membranes by droplet permeation, which can be quantified by permeate flux and regulated by feed flow rate. PTFE membrane has the best performance in structural stability and hydrophobic durability, and has more potential for engineering applications. This study provides a key basis for the selection and operation of hydrophobic membrane used for concentration of radioactive waste liquid by membrane distillation technology.
Americium-241(241Am)is an important transuranic nuclide with a half-life of 432.6 years,and is classified as a highly toxic radionuclide.Due to its rapid migration and potential for bioaccumulation along food chains in the environment,241Am poses potential threats to environmental radiation safety and human health.The concentration of 241Am in environmental water samples is extremely low,necessitating the collection of large-volume water samples to enhance detection sensitivity and enable the analysis and measurement of low-level 241Am.Pretreatment of large-volume water samples is a major challenge in chemical separation procedures.In this paper,50 L water samples were used for 241Am measurement,with systematic investigation of key factors such as coprecipitation methods and precipitate carrier amounts.After separation and purification,a suitable method was selected through systematic discussion.Finally,an analytical method for 241Am determination in environmental water samples using alpha spectrometry was established.In the process of water sample pretreatment,tap water samples(50 L)were used to compare the pre-concentration efficiencies of calcium-magnesium co-precipitation and manganese dioxide(MnO2)precipitation were compared,and the MnO2 precipitation method was selected for pre-concentration.Approximately 2.0 g of potassium permanganate(KMnO4)was added to form MnO2 precipitates,thereby enriching Am from the water matrix.In the separation and purification step,extraction chromatography was used to systematically investigate the effects of key parameters(i.e.,loading acidity and elution volume)on the chemical recovery yield.DGA resin was used to adsorb Am from the sample solution under 5-6 mol/L HNO3 conditions,followed by elution with 20 mL of 0.01 mol/L HCl-0.03 mol/L NaNO2 to achieve effective separation of Am from other interfering nuclides.For source preparation after separation and purification,0.1 mol/L(NH4)2SO4 was used as the electrolyte solution,and the pH was adjusted to approximately 2.Electrodeposition was carried out at a current density of 0.65-0.70 A/cm2 for about 60 minutes to prepare the measurement source for alpha spectrometry.Thus,an analytical method for 241Am determination in environmental water samples based on alpha spectrometry was established.The established analytical procedure was validated.The experimental results show that with a 48 h measurement time by alpha spectrometry,the overall chemical recovery yield of 243Am in the procedure is 83.4%,the decontamination factor for 242Pu exceeds 103,and the minimum detectable activity concentration of 241Am is determined to be 9.48×10-6 Bq/L.These results indicate that the established method is accurate and reliable.Finally,the method is applied to analyze 241Am in various environmental water samples,including surface water,groundwater,and river water,for monitoring radiation safety in the aquatic environment,meeting the requirements for routine analysis and monitoring of low-level 241Am in environmental water samples.
The safe and efficient immobilization of high-level liquid waste(HLLW) is a critical global challenge. Vitrification utilizing a continuous Joule-heated ceramic melter(JHCM) is currently the most effective technology for HLLW treatment. During operation, HLLW and glass-forming additives form a floating layer of unmelted batch materials, known as the cold cap, on the melt pool. The cold cap coverage rate significantly governs heat transfer dynamics, physical field distributions, and operational safety. Due to the extreme high-temperature and high-radiation environment inside the melter, directly obtaining internal physical field data is exceedingly difficult. Therefore, establishing an accurate predictive model for the melter’s performance is crucial for ensuring safe operation. In this study, a comprehensive three-dimensional numerical simulation was conducted using COMSOL Multiphysics software to systematically investigate the impact of different cold cap coverage rates(25%, 50%, 75%, and 90%) on the JHCM’s melting efficiency and safety. The computational model coupled fluid dynamics, heat transfer, electric fields, and Joule heat, utilizing approximately 200 000 mesh elements. The molten glass flow was treated as an incompressible Newtonian fluid in a laminar regime, whereas the plenum gas flow was simulated using a turbulent model. The simulation results demonstrate that as the cold cap coverage increases from 25% to 90%, the average temperature within the melt pool rises from 1127 ℃ to 1241 ℃, an increase of approximately 10%. Conversely, the expanded cold cap effectively insulates the upper space, causing the average temperature in the plenum to drop drastically from 897 ℃ to 635 ℃(a 30% reduction), and the exhaust gas pipeline temperature to decrease from 746 ℃ to 501 ℃. These variations profoundly influence electrical and thermal distributions. Under a low coverage mode, the distributions of current density and Joule heat within the molten glass are notably heterogeneous. Excessive local current density concentrates near the electrodes(exceeding 5000 A/m2), significantly exacerbating the risk of electrode ablation. In contrast, under high coverage conditions, the Joule heat distribution homogenizes remarkably, with regions exceeding 80 kW/m3 increasing to 54% of the pool volume. Energy distribution analysis shows that at 90% coverage, thermal energy dissipated into the plenum diminishes to 16.0 kW, compared to 73.5 kW at 25% coverage. Accordingly, heat flux from the melt pool to the cold cap escalates from 11.5 to 22.8 kW/m2. Consequently, the comprehensive melting efficiency of the melter undergoes a near-twofold enhancement, surging from 275.4 to 545.3 kg/(m2•d). In conclusion, maintaining high cold cap coverage optimizes internal multiphysics fields, mitigates electrode burnout risks, and maximizes vitrification efficiency.
Nuclear energy, as a safe, low-carbon, stable, and efficient clean energy source, demonstrates broad development prospects in global energy transition and climate change mitigation. Uranium is a key raw material for nuclear energy, and its demand continues to grow with the rapid development of the nuclear industry. Seawater contains abundant uranium reserves, nearly a thousand times that of terrestrial uranium. Therefore, uranium extraction from seawater is crucial for the sustainable development of nuclear energy. Currently, there are various methods for seawater uranium extraction, such as adsorption, ion exchange, membrane separation, photocatalysis, and electrochemical methods. However, each method has its own advantages and disadvantages. Combining different methods to leverage their strengths can potentially improve uranium extraction efficiency. In this study, a porous composite aerogel material, PAO/g-C3N4, was prepared by first combining adsorbent and photocatalytic materials into a hydrogel through molecular crosslinking, followed by freeze-drying. The material was characterized by SEM, FTIR, XRD, and other techniques. Batch experiments were conducted to investigate the effects of solid-to-liquid ratio, pH, ionic strength, contact time, and initial concentration on adsorption performance. The results show that the prepared aerogel material has a porous network structure, which exposes more active sites and increases the contact opportunities between the material and uranyl ions. Additionally, the composite aerogel PAO/g-C3N4 retains the original properties of its components. Adsorption experiments indicate that both PAO/g-C3N4 and PAO follow pseudo-second-order kinetics, with optimal uranium adsorption under near-neutral conditions. The uranium adsorption capacity of PAO fits the Langmuir adsorption model, suggesting monolayer chemical adsorption. In contrast, the uranium extraction capacity of PAO/g-C3N4 does not reach saturation within the studied concentration range, indicating its excellent uranium extraction capability. The uranium extraction performance of both PAO/g-C3N4 and PAO is influenced by pH and ionic strength, suggesting that inner-sphere and outer-sphere complexation may be the adsorption mechanism. Moreover, PAO/g-C3N4 exhibits good reusability and ion selectivity. In simulated seawater with extremely low uranium concentration, PAO/g-C3N4 shows higher uranium extraction than PAO. Mechanistic analysis reveals that PAO relies solely on adsorption, while PAO/g-C3N4 combines both adsorption and photocatalysis. The amidoxime groups in PAO primarily coordinate with uranyl ions, while g-C3N4 acts as a photocatalyst. Due to the presence of dissolved oxygen during the experiment, the photocatalytic product of g-C3N4 is uranyl peroxide dihydrate. Under the combined effects of adsorption and photocatalysis, PAO/g-C3N4 achieves efficient uranium extraction, demonstrating potential for practical applications. Furthermore, this study provides a new approach for the synergistic application of adsorption and photocatalysis.
The efficient conversion of metals into nitrides and oxides is one of the key steps in nuclear fuel processing. In this work, metallic cerium(Ce) was used as a surrogate material. The nitridation reaction of metallic Ce was achieved in a mixed atmosphere of low-concentration water vapor(water vapor partial pressure ≤ 5.0 kPa) and nitrogen(N2). The effects of water vapor partial pressure(0.5, 1.0, 2.5 and 5.0 kPa) and reaction temperature(310, 330, 350 and 370 ℃) on the nitridation reaction rate of metallic Ce were investigated. According to the pressure change during the reaction process and the characterization results of X-ray diffraction(XRD), it is found that metallic Ce reacts with N2 in the presence of low-concentration water vapor and is completely converted to cerium nitride(CeN). Increasing the partial pressure of water vapor and the reaction temperature can accelerate the nitridation reaction rate of metallic Ce. When liquid water is used instead of water vapor, metallic Ce can still undergo the nitridation reaction and be converted into CeN, further simplifying the nitridation reaction process. In addition, through the characterization by XRD, Raman spectroscopy and X-ray photoelectron spectroscopy, it is demonstrated that the CeN product is rapidly oxidized and converted into high-purity cerium dioxide(CeO2) in an air atmosphere. This work provides new insights into the mechanisms and processes of metal nitridation-oxidation reactions.
210Pb is an important long-lived radionuclide in the 238U decay series with a half-life of 22.23 years,which shows high radiotoxicity and chemical toxicity and widely exists in the atmosphere,soil,water and biological organisms.210Pb originates from both natural decay of uranium and anthropogenic activities such as coal combustion and phosphate fertilizer production,and it can be concentrated through the food chain and cause continuous internal radiation exposure to humans via diet,so accurate measurement of 210Pb in biological samples is essential for radiation exposure assessment,food safety control and environmental ecological research.However,the determination of 210Pb in biological samples faces many difficulties including complex matrix,serious interference from coexisting elements and very low activity concentration,traditional methods such as γ-spectrometry,α-spectrometry and liquid scintillation counting have obvious limitations in sensitivity,selectivity or analysis time,and the crown ether resins used in standard methods are expensive and provide unsatisfactory recovery in complex biological matrices,therefore this study aims to establish a low-cost,stable and reliable method for the determination of trace 210Pb in biological samples.In this work,a new method for 210Pb determination was developed using anion exchange resin separation combined with low background β-counter method,a pretreatment procedure of HNO3-H2O2 digestion followed by ashing and leaching was adopted,samples were digested with concentrated nitric acid and hydrogen peroxide,evaporated to dryness and then ashed at 450℃for 2 h,the residue was leached with 1 mol/L hydrochloric acid and filtered,the filtrate was passed through a 201×7(717)strong-base anion exchange resin column pre-equilibrated with 1 mol/L HCl,the column height-to-diameter ratio was optimized to 15∶2-20∶2 and the flow rate was controlled at 1 mL/min,impurities were rinsed with 1 mol/L HCl and deionized water and lead was eluted with 80 mL deionized water,the eluate was concentrated and converted to lead sulfate precipitate,the precipitate was filtered,washed and dried at 105℃to constant weight,the sample source was kept for more than 30 days to reach radioactive equilibrium between 210Pb and 210Bi,and the β-counting rate of 210Bi was measured with a low background α/β-counter to calculate the activity concentration of 210Pb.The method exhibits excellent performance and is verified by seven laboratories,the detection limit is 0.9 mBq/g(ash)which meets the requirements of relevant environmental monitoring standards,the precision is satisfactory with the relative standard deviation(sr)below 10%(n=6),the trueness is high with relative errors ranging from-17%to 17%and mean relative errors from-3.6%to 0.73%,the optimized separation process effectively removes matrix ions such as Ca2+,Mg2+,Fe3+,and radioactive interferences such as 90Sr,the 201×7(717)resin provides high selectivity,low cost and stable recovery for 210Pb separation,and inter-laboratory validation confirms that the method has good repeatability and reproducibility.This method is simple to operate and has strong anti-interference ability,it is suitable for rapid separation and accurate determination of trace 210Pb in various biological samples,provides reliable technical support for radiation environmental monitoring and food safety assessment,and can serve as a useful reference for the formulation of relevant standard analytical methods.
Deuterated water(D2O), as an important chemical raw material, plays an irreplaceable role in the nuclear energy industry, particularly as a moderator and coolant in nuclear reactors. Water distillation is regarded as an effective method for dehydrogenation and deuterium enrichment, and its separation efficiency directly determines the purity of D2O. However, due to the extremely small vapor pressure difference between D2O and H2O, highly efficient gas-liquid contact and mass transfer within the column are required in the D2O-H2O distillation process. Among the various factors, packing performance is identified as a key determinant of mass transfer efficiency. The wettability of the packing surface is known to significantly influence liquid film spreading, interfacial renewal, and the effective mass transfer area. Although extensive studies on packing materials have been reported, the mechanism by which packing wettability regulates mass transfer behavior in D2O-H2O isotope distillation remains insufficiently understand, particularly when dynamic interfacial phenomena are taken into account. In this study, a combined experimental and theoretical modeling approach is employed to systematically investigate the effect of packing surface wettability on the mass transfer performance of D2O-H2O distillation. Copper-based packings with different hydrophilicities are prepared via surface modification, and their surface morphology, crystal structure, elemental distribution, wettability, and long-term stability are characterized. Under total reflux conditions, the height equivalent to a theoretical plate(HETP) is adopted as the primary evaluation metric to assess mass transfer performance. Furthermore, the influence of the packing surface contact angle and gas velocity on HETP and pressure drop is analyzed. An HETP prediction model associated with mass transfer is established based on the surface renewal theory. By introducing a wettability parameter, the model couples packing surface properties with liquid film renewal behavior, thereby elucidating the influence of dynamic gas-liquid interfacial renewal on the mass transfer process. The results show that enhancing the hydrophilicity of the packing improves interfacial renewal and significantly reduces the HETP value. A moderate increase in gas velocity further enhances mass transfer efficiency without a noticeable rise in pressure drop, indicating a synergistic effect between wettability and operating conditions. The proposed HETP prediction model shows excellent agreement with experimental data, with a coefficient of correlation(R2) of 0.996 and relative errors controlled within 20%, indicating good predictive accuracy and applicability. These findings demonstrate that, compared with the conventional two-film theory, the surface renewal model provides a more realistic description of the dynamic mass transfer behavior in D2O-H2O distillation. This study provides a theoretical basis for packing design and process optimization in the D2O-H2O distillation system and offers valuable guidance for isotope enrichment and the design of high-efficiency distillation columns.
The catalytic oxidation treatment of radioactive trioctylamine-xylene(V/V=1/4) organic waste liquid containing transuranic nuclides(total α activity: 108 Bq) using commercially available nano-MnO2 as the catalyst was investigated in the present work. The process aims to convert organic components into non-dispersible solid residues while minimizing radionuclide release. The simulated experiments were conducted to validate the process over 90 hours of continuous operation. Results demonstrate that the trioctylamine-xylene mixture is predominantly decomposed into carbon dioxide, water, and trace small-molecule organics, achieving an inorganic conversion rate exceeding 95%. Key reaction parameters, including a reactor temperature of (190±20) ℃, stirring rate of 30 r/min, and feed rate of 87 mL/h, are optimized to ensure stable operation. Post-reaction analysis reveals that residual carbon content in the catalyst increased from 0.044% to 2.08%, indicating minor carbon deposition. Tail gas analysis detects volatile organic compounds(TVOC: 925.1 mg/m3), predominantly xylene(660.60 mg/m3), alongside trace benzene, toluene, and olefins, suggesting partial catalytic oxidation and chemical reforming pathways. Notably, trioctylamine exhibits higher catalytic degradation efficiency compared to xylene, likely due to its stronger polarity. Nitrogen oxides(NO: 0.05 mg/m3, NO2: 0.52 mg/m3) in the tail gas are minimal, implying nitrogen retention in solid residues or conversion to N2. Subsequent experiments in hot cell were successfully carried out to treat radioactive waste liquid under optimized conditions. Post-treatment radionuclide balance calculations reveal that ＞98% of transuranic nuclides are retained in catalyst residues and reactor internals. Less than 0.001% of nuclides are released via gaseous pathways, confirming effective containment. Solid residues constitute 96.86% of the original activity. The catalytic oxidation system demonstrated robust performance in converting liquid organic waste into stable solid residues under mild conditions(190 ℃, atmospheric pressure), avoiding secondary pollution risks compared with high-temperature incineration or corrosive supercritical oxidation. This work validates catalytic oxidation as a viable method for treating radioactive organic liquids, particularly for small-batch operations. The process achieves high radionuclide immobilization while enabling safe gas-phase discharge, offering significant advantages over conventional cementation(e.g., volume expansion, leaching issues) and glass vitrification(incompatible with organics). Further optimization of catalyst formulations and tail gas treatment can enhance decomposition efficiency for complex organic matrices.
The efficient separation and recovery of neptunium(Np), one of the most chemically complex actinides in the nuclear fuel cycle, represent a critical challenge in spent nuclear fuel reprocessing and advanced fuel-cycle management. Effective control of Np behavior is essential not only for improving the sustainability and economics of nuclear energy systems, but also for minimizing the long-term radiotoxicity of radioactive waste and supporting the implementation of a closed nuclear fuel cycle. Among the known neptunium isotopes, 237Np is of particular importance because of its long half-life(2.1×106 years), high radiotoxicity, and potential application as a precursor for the production of 238Pu, which is widely used as a heat source in radioisotope thermoelectric generators. However, the extraction and separation of Np remain exceptionally difficult owing to its complicated electronic structure and the coexistence of multiple oxidation states, including Np(Ⅲ), Np(Ⅳ), Np(Ⅴ), Np(Ⅵ), and Np(Ⅶ), whose stability and chemical behavior strongly depend on solution composition, acidity, redox conditions, and coordinating ligands. This review provides a comprehensive overview of recent advances in Np extraction and separation technologies for spent nuclear fuel reprocessing. The fundamental chemistry of neptunium, particularly its redox properties and valence-state transformations, is first summarized to establish the theoretical basis for separation processes. Subsequently, the major extraction and recovery approaches, including precipitation, solid-phase adsorption, membrane-based separation, and solvent extraction, are systematically reviewed and compared with respect to their separation mechanisms, process efficiencies, advantages, and limitations. Particular attention is given to solvent-extraction technologies because of their dominant role in industrial reprocessing. Recent developments involving tributyl phosphate, monoamides, diglycolamides, triazine-based ligands, and other functional extractants are discussed in detail. In addition, strategies based on valence-state control, including chemical, electrochemical, and photochemical methods, are critically evaluated because precise manipulation of Np oxidation states is often the key factor determining its distribution and recovery behavior. The review further summarizes representative advanced reprocessing flowsheets developed worldwide, including UREX+(NPEX), PARC, APOR, and other Np-oriented separation schemes. Their process configurations, separation principles, Np routing characteristics, and engineering performance are analyzed and compared. Particular emphasis is placed on the use of salt-free organic reductants, selective complexants, and integrated process-control strategies that enable more efficient and environmentally benign Np management. The major technical barriers to industrial implementation are also discussed, including the complexity of Np redox chemistry, limitations in extractant selectivity and radiation stability, process integration challenges, and the generation of secondary waste streams. Finally, future research directions are proposed, including deeper investigation of fundamental Np chemistry in diverse process environments, development of highly selective and radiation-resistant functional materials, advancement of intelligent process monitoring and control technologies, and integration of innovative separation strategies into next-generation reprocessing flowsheets. This review aims to provide a comprehensive theoretical and technical reference for improving Np recovery efficiency, enhancing nuclear fuel cycle safety, and promoting the realization of sustainable closed nuclear fuel cycle systems.
With the rapid development of medical radioactive isotopes, nuclear medicine has played an increasingly important role in the diagnosis and treatment of various diseases. Medical isotopes such as iodine-125(125I), iodine-131(131I), strontium-89(89Sr), technetium-99m(99Tcm) have been widely utilized in clinical imaging examinations, targeted radionuclide therapies, and other related fields. While these isotopes make significantly contributions to various aspects of nuclear medical practices, their application inevitably produces substantial volumes of radioactive medical liquid waste. Before these wastes can be discharged, they generally require a period of decay storage to reduce their radioactivity to levels compliant with the discharge limits set by relevant national regulations and standards. In addition, the complex radionuclide species, wide concentration distribution, and diverse chemical forms in the waste significantly elevate the difficulty of separation and treatment, it imposes higher requirements on the selectivity and stability of the treatment materials. Consequently, the safe and efficient treatment and disposal of such waste has become a critical and urgent issue requiring immediate attention. At present, to effectively reduce environmental risks and ensure public safety, a diversified technical system has been established for the treatment and disposal of medical radioactive liquid waste. This system primarily encompasses technologies such as the decay pool method, membrane treatment, chemical separation and precipitation, evaporation concentration, as well as adsorption and ion exchange techniques. In contrast, the adsorption approach boasts the advantages of high adsorption efficiency, straight forward maintenance and operation, cost-effectiveness, minimal energy consumption and scalability for large-scale applications, has become the most widely employed method for radionuclide separation and removal. In the last years, in order to remove radioactive nuclides from waste liquids, various high-performance adsorbent materials have been developed and reported, including zeolites, activated carbon, layered materials, metal-organic frameworks(MOFs), covalent organic frameworks(COF), porous organic polymers(POPs), ion-exchange resins, and nanocomposite materials. This paper provides a comprehensive review of the latest research progress and current applications in the treatment and disposal technologies for medical radioactive liquid waste, with a particular focus on the development and optimization of radionuclide removal materials. Furthermore, considering the current status of nuclear technology applications and waste management needs in China, the future development directions of medical liquid waste treatment technologies are discussed. It aims to provide theoretical references and practical guidance for the application of related materials in the treatment of medical radioactive waste.
Uranium-niobium-titanium alloy has important application value in the nuclear energy field due to its excellent corrosion resistance and mechanical properties. Accurate and efficient determination of niobium(Nb) and titanium(Ti) content in uranium-niobium-titanium(U-Nb-Ti) alloys is critical for ensuring the performance of nuclear materials. Conventional analytical methods, such as gravimetry, spectrophotometry, and inductively coupled plasma atomic emission spectroscopy(ICP-AES), are hindered by complex procedures, prolonged analysis time, and matrix interferences. This study introduces a novel methodology based on wavelength-dispersive X-ray fluorescence spectrometry(WDXRF) for simultaneous quantification of Nb(w=1.0%-7.0%) and Ti(w=0.2%-0.9%) in U-Nb-Ti alloys, addressing the limitations of traditional techniques while enhancing analytical efficiency and reliability. The alloy samples were dissolved in a mixture of nitric acid(HNO3) and hydrofluoric acid(HF) under controlled heating. To prevent hydrolysis of Nb and Ti during solution stabilization, saturated citric acid was added prior to volumetric dilution, ensuring homogeneity and long-term stability of the solution. The resulting solution was immobilized on polyester filter paper to form uniform thin-film specimens, effectively minimizing matrix effects. Calibration standards were prepared using certified reference solutions of U, Nb, and Ti, with critical instrumental parameters—including crystal selection, detector configuration, and operating conditions optimized to establish a robust linear relationship between elemental mass ratios(Nb/U, Ti/U) and their characteristic X-ray intensity ratios(Nb Kα/U Lβ2, Ti Kα/U Lβ2). Calibration curves exhibit excellent linearity, with correlation coefficients(r2) exceeding 0.99. Method validation confirms that the measured values agree well with those obtained by ICP-AES and chemically prepared standards, demonstrating excellent consistency. Precision tests on six parallel samples yield relative standard deviations(sr) of ≤1% for Nb and ≤2% for Ti, confirming superior method reproducibility. Long-term stability assessments over 30 days revealed sr below 2%, underscoring the effectiveness of citric acid in stabilizing Nb/Ti solutions and maintaining the integrity of thin-film specimens. The integration of citric acid not only suppresses hydrolysis but also enhances measurement consistency, particularly for low-concentration Ti(w=0.2%-0.9%). This WDXRF-based method achieves high instrumental efficiency, with a measurement time of only 50 seconds per sample, eliminating time-consuming separation steps and offering a practical solution for industrial quality control of U-based alloys. The thin-film preparation technique combined with citric acid stabilization effectively mitigates matrix interference and hydrolysis-related challenges. Future research should focus on extending this approach to major element analysis in other alloy systems or complex matrices, leveraging its adaptability to advance nuclear material characterization technologies.
Under the current background of waste resource recycling utilization, people have begun to actively explore the sustainable development path of nuclear fuel recycling. The innovative application of ammonium phosphomolybdate(AMP) in the field of nuclear wastewater treatment is one of the research hotspots in the sustainable development of nuclear energy, which mainly involves the separation, purification and concentration of high heat release fission product 137Cs from radioactive. However, AMP has small crystal grains, which makes it difficult to separate from waste liquid and prone to clogging adsorption columns when used. This limits its large-scale industrial application. In recent years, with in-depth research on Cs+ adsorbents, various AMP-based adsorbents have emerged, which are prepared by optimizing the synthesis route to produce larger particle sizes of AMP or by loading AMP onto suitable carriers. These innovations are expected to solve the aforementioned challenges. This article summarizes various methods for the preparation of large-size AMP, as well as the composite methods of AMP with organic carriers(polyacrylonitrile, polymethyl methacrylate, and alginate) and inorganic carriers(porous silica, metal-organic framework, magnetic materials and alumina). It reviews the research progress on the adsorption of Cs+ from aqueous solutions by large-size AMP and AMP-loaded composite adsorbents in recent years. By considering the adsorption performance of each adsorbent, including adsorption capacity, distribution coefficient, acid and alkali resistance, radiation resistance, and recyclability, the unique advantages of coarse-grained AMP and various carriers are summarized, in the hope of providing a reference for the application of AMP in the efficient separation of Cs from nuclear wastewater.
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.
The decommissioning management process of nuclear facilities requires rapid classification and detection of soil radionuclides and contamination levels, of which 90Sr is one of the focus nuclides. Both 90Sr and its daughter 90Y emit only β-rays, which cannot be measured directly due to the limited capacity of β-ray penetration. The traditional method is sampling and separation. 90Sr is separated from the soil by radiochemical process, and then 90Sr activity is measured. This process is time-consuming and labor-intensive, cannot rapidly analyze the 90Sr content, and in addition, generates a large amount of secondary radioactive waste. In order to solve the problem of direct measurement of 90Sr in environmental soils, a stacked plastic scintillation fiber(SCiFi) array detector with a 5-layer structure was developed. This detector utilizes the characteristics of the difference in the energy of different β-rays as well as the difference in the interactions of different rays with matter, and combines a specific detector structure and signal analysis logic, to realize the detection of high-energy β-rays of 90Y from γ-, β-, and cosmic-ray radiation environments, thus realizing the direct measurement of 90Sr. The detector consists of 4 layers of 800 plastic scintillation optical fibers of 1 mm squar, and every 25 fibers are divided into a group to connect a piece of silicon photomultiplier device(SiPM) of 6 mm×6 mm, which converts the scintillation fluorescence into electrical signals, and the logical relationship between signals is processed by self-developed integrated circuit based on field programmable gate array(FPGA). For detectors with large-area flat plate structure, the variation of detection efficiency at different position is an important parameter. The detection efficiency was measured in different areas. And the minimum deviation of detection efficiency at different positions along the optical fiber length of 40 cm is 2.7%, which indicates that the effective length of the scintillation optical fiber is suitable, and detectors with larger sizes can be processed. In addition, the use of SiPM optimizes the detector structure and allows the fabrication of smaller detectors. The key performance parameters of the detector were tested. The effective detection area of the detector is 800 cm2, the detection efficiency of 90Sr in soil is 1.2%, the lower limit of detection is 0.2 Bq/g(5 min), and several real contaminated soils were measured with the same 90Sr content, which meets the demand for direct measurement of soil 90Sr content.
Due to the specific characteristics of high linear energy transfer(80-100 keV/μm), short penetration distance(40-100 μm), strong cytotoxic and few reverse effects, α radionuclides has a promising prospect for the therapy of micro-tumors, primary and micro-metastatic cancers. As the parent nuclide of the alpha-emitting radioisotope 212Bi, 212Pb has been attracted much attention in the application of targeted alpha therapy. Compared with 225Ac, 212Pb only involves one alpha decay. This ensures that therapeutic energy is delivered precisely and reduces the damage to the surrounding healthy tissues. In addition, Pb has a strong coordination ability and is more likely to combine with small molecule ligands to form stable targeted drugs. Thus, in recent years, researchers have prepared 212Pb radiopharmaceuticals by combining 212Pb with monoclonal antibodies, peptides, nanoparticles, etc. This paper summarizes the properties and preparation techniques of 212Pb nuclides, and the properties of chelators between 212Pb and targeted ligands. Three different ways of preparation technologies for separating and extracting of 212Pb are described: through 232Th or 232U, reactor irradiation of 226Ra. In contrast, there are some advantages to produce 212Pb by the third method. The extraction efficiency of 228Th can be increased and the loss of 226Ra is extremely low during the preparation process. Then, this article describes the 212Pb-complexes with DOTA, NOTA, TCMC, p-SCN-Bn-TCMC. The research progress and current clinical application of 212Pb radiopharmaceuticals, such as 212Pb-DOTAMTATE, 212Pb-ADVC001, 212Pb-DOTAM-GRPR1, 212Pb-VMT-α-NET, are also discussed. In addition, 203Pb, as an easily accessible diagnostic nuclide, can be used in combination to provide accurate data for the 212Pb treatment, which can further increase the application of 212Pb radiopharmaceuticals in the clinical. The rapid development of 212Pb radiopharmaceuticals has highlighted their potential as a form of radiotherapy and their remarkable effects in treating a variety of diseases. It is expected to be a support for providing references for the researches and applications in the treatment of malignant tumors in the future. Meanwhile, it should be noted that a large amount of research is still needed before it can be used for human treatment in the hospital, including production and transportation of 212Pb nuclides, the recoil phenomenon of decayed nuclides, the visual monitoring of drug distribution during treatment, the estimation of radiation doses, the assessment of toxicity, and the setting of standard doses, etc. Although large-scale clinical research is still needed, we should believe that 212Pb radiopharmaceuticals can play a significant role in the future with the continuous development of nuclear and medical technologies.
Using Baiyun Ebo fluorite tailings and Baogang steel blast furnace slag as the main raw materials, a certain amount of quartz sand and borax were added to prepare glass-ceramics. The influence of raw material ratio and preparation process on the radioactivity of glass-ceramics was studied. Using a low background multi-channel gamma spectrometer, radioactive nuclide detection was carried out on various raw materials and glass-ceramics products. Based on the radioactive indicators of each raw material and the basic glass formula, the radioactive indicators of glass-ceramics products were predicted by calculation according to the principle of linear superposition. The research results show that during the process of gradually changing the mass ratio of fluorite tailings to blast furnace slag from 1∶9 to 9∶1 in the raw materials, with the increase of fluorite tailings ratio, the internal irradiation index of glass-ceramics shows a gradually decreasing trend, and the external irradiation index shows a gradually increasing trend. The preparation process basically does not affect the internal and external irradiation index of glass-ceramics; the relative errors between the predicted and detected values of the internal and external irradiation indices are controlled within 16% and 10%, respectively. The measurement relative error ranges of the internal and external irradiation indices are 9% to 13% and 7% to 11%, respectively, which can be used as an effective means of predicting product radioactivity. When the mass ratio of Baiyun Ebo fluorite tailings to Baogang blast furnace slag is 1∶9, 2∶8, and 3∶7, the glass-ceramic meets the B-class decoration material standard. When the mass ratio of Baiyun Ebo fluorite tailings to Baogang steel blast furnace slag is 4∶6, 5∶5, and 6∶4, the glass-ceramic meets the C-class decoration material standard. However, the radioactive isotopes of 7∶3, 8∶2, and 9∶1 glass-ceramic exceed the standard and cannot be used as building decoration materials, providing basic information and theoretical basis for the green and high-value utilization of solid waste from Baiyun Ebo mine selection and smelting.
