Search Results (54)

Liquid metal cathodes, particularly liquid cadmium (Cd), are widely used in molten salt electrorefining and pyrochemical reprocessing of spent nuclear fuel due to their high electrical conductivity, strong affinity for actinides, and favorable alloying characteristics. During electrorefining, reduced metal species enter the liquid Cd phase and may exist as dissolved atoms, liquid alloys, or intermetallic compounds, all of which significantly influence deposition behavior, separation selectivity, and cathode performance. Although numerous experimental and theoretical studies have investigated metal solubility, alloy formation, and diffusion in liquid Cd systems, the current understanding remains fragmented. Thermodynamic phase behavior and mass transport kinetics are often discussed separately, and reported diffusion data show considerable discrepancies owing to variations in experimental techniques and interpretations. In addition, the relationship between phase stability, diffusion mechanisms, and electrochemical conditions in practical electrorefining environments has not yet been systematically clarified. This review aims to present an integrated thermodynamic–kinetic perspective on the behavior of metals in liquid Cd cathodes. Recent progress in dissolution behavior, alloy phase formation, and diffusion-controlled transport processes is critically summarized. The differences in solubility and precipitation behavior of actinides, rare-earth elements, and selected transition metals are analyzed in relation to binary phase diagrams and thermodynamic stability. Furthermore, experimental methods for determining diffusion coefficients, including capillary techniques and electrochemical approaches, are comparatively evaluated. By correlating thermodynamic phase stability with diffusion-driven mass transport, this work provides a coherent framework for understanding metal behavior in liquid Cd cathodes and offers insights for optimizing molten salt electrorefining and advanced nuclear fuel cycle technologies. Full article

The pyrometallurgical reprocessing of spent fuel developed by the United States is currently one of the most promising nuclear fuel reprocessing methods. The electroreduction, electrolytic refining, and electrodeposition processes involve electrochemical research in high-temperature molten chloride systems. In recent years, much progress has been made in simulating and studying molten-salt systems from a microscopic perspective using the first-principles molecular dynamics (FPMD) simulation technique. Using this method for simulation calculations is more conducive to analyzing the microscopic action mechanism and microscopic mechanism in the system from the atomic level and explaining the internal reasons for various electrochemical behaviors and phenomena. This opens up a new path for the study of molten-salt electrochemical systems. However, there are still a few systematic reviews of simulating work in first-principles computation. Therefore, this work summarizes the theoretical calculation work on molten-salt electrochemical systems of recent years, focusing on the research progress in computational aspects such as coordination properties, physical properties, and electrode behavior, which has good guiding value for the application of FPMD in molten-salt electrochemistry. Full article

Some of nuclear power’s primary detractors are the unique environmental challenges and impacts of radioactive wastes generated during fuel cycle operations. Key benefits of spent fuel reprocessing (SFR) are reductions in primary high active waste (HAW) masses, volumes, and lengths of radiotoxicity at the expense of secondary waste generation and high capital and operational costs. By employing advanced waste management and resource recovery concepts in SFR beyond the existing standard PUREX process, such as minor actinide and fission product partitioning, these challenges could be mitigated, alongside further reductions in HAW volumes, masses, and duration of radiotoxicity. This work assesses various current and proposed SFR and fuel cycle options as base cases, with further options for fission product partitioning of the high heat radionuclides (HHRs), rare earths, and platinum group metals investigated. A focus on primary waste outputs and the additional energy that could be generated by the reprocessing of high-burnup PWR fuel from Gen III(+) reactors using a simple fuel cycle model is used; the effects of 5- and 10-year spent fuel cooling times before reprocessing are explored. We demonstrate that longer cooling times are preferable in all cases except where short-lived isotope recovery may be desired, and that the partitioning of high-heat fission products (Cs and Sr) could allow for the reclassification of traditional raffinates to intermediate level waste. Highly active waste volume reductions approaching 50% vs. PUREX raffinate could be achieved in single-target partitioning of the inactive and low-activity rare earth elements, and the need for geological disposal could potentially be mitigated completely if HHRs are separated and utilised. Full article

Nuclear energy can help stop climate change by generating large amounts of emission-free electricity. Nuclear reactor designs are continually being developed to be more fuel efficient, safer, easier to construct, and to produce less nuclear waste. The term advanced nuclear reactors refers either to Generation III+ and Generation IV or small modular reactors. Every reactor is associated with the nuclear fuel cycle that must be economically viable and competitive. An important matter is optimization of fissile materials used in reactor and/or reprocessing of spent fuel and reuse. Currently operating reactors use the open cycle or partially closed cycle. Generation IV reactors are intended to play a significant role in reaching a fully closed cycle. At the same time, we can observe the growing interest in development of small modular reactors worldwide. SMRs can adopt either fuel cycle; they can be flexible depending on their design and fuel type. Spent nuclear fuel management should be an integral part of the development of new reactors. The proper management methods of the radioactive waste and spent fuel should be considered at an early stage of construction. The aim of this paper is to highlight the challenges related to reprocessing of new forms of nuclear fuel. Full article

N,O-donor hybrid heterocyclic extractants have great potential for separation of actinides from lanthanides in spent nuclear fuel reprocessing processes. We demonstrate that this type of reagents can be used for primary concentration of actinides contained in eudialyte, a promising mineral containing a heavy group of lanthanides. With respect to lanthanide ions, the efficiency of their extraction decreases in the series L3 >> L1 > L2, and the extraction of actinides decreases in the series L1 ≈ L3 >> L2. For the extractant L2 based on 2,2′-bipyridine-6,6′-dicarboxylic acid diamide, the efficiency of lanthanide purification from U, Th exceeds 50. The structure and stereochemical features of the ligands do not have a significant effect on the composition of the formed complexes. The solvation numbers are close to 1 for all range f-elements studied, except for thorium, which indicates the predominant formation of complexes with the composition ratio of 1:1. The solvation numbers 1.4–1.5 are observed for thorium(IV), and the established values indicate the formation of a mixture of complexes with the composition ratios of 1:1 and 2:1. Full article

Radioactive metal oxides are highly radioactive, hygroscopic spent fuel reprocessing products generally stored in container-sealed dry storage. During the storage process of metal oxides, a large amount of heat is generated due to radioactive decay, and helium is produced by α-decay, which leads to an increase in the temperature and pressure of the storage container. In order to ensure the safety of the radioactive metal oxides in the long-term storage process, computational fluid dynamics simulations are used to investigate the effects of storage conditions on the temperature and pressure of the container. Based on a large amount of simulated temperature data under different storage conditions, a power function is used to construct a mathematical model of ventilation speed, ventilation temperature, stack density, loading volume, heating power, water content, and cumulative helium mass versus metal oxide temperature to obtain a safe, reliable, and economical storage method. The results show that reducing the loading volume and increasing the density of metal oxides, increasing the ventilation speed, and lowering the ventilation temperature are beneficial to the heat transfer and cooling in the dry storage process; increasing the density of metal oxides and lowering the water content of metal oxides and increasing the ventilation temperature and speed are beneficial to avoid the high pressure inside the container. Based on the optimized storage conditions, the temperature peak in the storage process occurs near 25 years, and its temperature reaches 527.6 K. The mathematical model of storage temperature constructed in this study has high computational accuracy, and the maximum relative error of storage temperature is less than 1.80%. Full article

The PUREX process is a key technology for spent fuel reprocessing, designed to selectively recover uranium and plutonium mainly through multiple chemical separation stages, minimizing high-level waste. Acetohydroxamic acid (AHA) enhances selectivity in this process but decomposes into acetic acid (HAc), which disrupts chemical equilibrium and reduces extraction efficiency. This study examines the extraction and separation of nitric acid (HNO₃) and HAc using 30% tributyl phosphate in organic kerosene (TBP-OK) under various conditions. Results show that 30%TBP-OK preferentially extracts HAc over HNO₃, especially in the low acid concentration range (HNO₃ < 1 mol/L, HAc < 0.2 mol/L). The selectivity coefficient drops from 3.05 in a 0.5 mol/L HNO₃-0.1 mol/L HAc system to 2.18 in a 1 mol/L HNO₃-0.2 mol/L HAc system. TBP forms stable 1:1 complexes with both acids, with equilibrium constants around 0.85 under typical conditions. Increasing TBP concentration enhances HNO₃ extraction, while phase ratio adjustments improve HAc separation. A 16-stage countercurrent extraction simulation confirms that optimizing these factors effectively separates HNO₃ and HAc, offering theoretical and technical support for refining the PUREX process. Full article

The choice of efficient methods for the immobilization of high-level waste (HLW) resulting from the reprocessing of spent nuclear fuel (SNF) is an important scientific and practical task. The current policy of managing HLW within a closed nuclear fuel cycle envisages its vitrification into borosilicate (B-Si) or alumina–phosphate (Al-P) glasses. These wasteforms have rather limited waste loading and can potentially impair their retaining properties on devitrification. The optimal solution for HLW immobilization could be separating radionuclides into groups using dedicated capacious durable matrices. The phases of the Nd₂O₃–ZrO₂–TiO₂ system in this respect are promising hosts for the REE (rare earth elements: Nd, Ce, La, Pr, Sm, Gd, Y) –MA (MA: Am, Cm) fraction of HLW. In this manuscript, we present data on the composition of the samples analyzed, their durability in hot water, their behavior under irradiation, and their industrial manufacturing methods. Full article

Iodine is one of the key elements that must be removed from the off-gas systems of nuclear fuel reprocessing. This study systematically investigates the iodine vapor adsorption performance of the metal–organic framework (MOF) material HKUST-1(1-(2-methyl-4-(2-oxopyrrolidin-1-yl)phenyl)-3-morpholino-5,6-dihydropyridin-2(1H)-one), with particle sizes of 100 nm and 20 μm. HKUST-1 samples with varying particle sizes were synthesized via a hydrothermal method. The experimental results show that the 20 μm HKUST-1 exhibits superior crystallinity, a more intact pore structure, and a higher iodine adsorption capacity, reaching 700 mg/g, which is significantly greater than the 300 mg/g capacity of the 100 nm HKUST-1. Kinetic analysis reveals that the adsorption process follows the pseudo-second-order model, with physical adsorption as the predominant mechanism, where iodine molecules are accommodated within the pores. FTIR and XRD further confirm the structural stability of the HKUST-1 framework after iodine adsorption. However, desorption experiments show that iodine molecules are easily volatilized into the air, with a 20% weight loss observed within 10 h and a color change from black to green. The results provide experimental evidence for optimizing the application of HKUST-1 materials in iodine capture and suggest that material modification could enhance the long-term stability of iodine fixation. Full article

Fast-neutron reactors are an important representative of Generation IV nuclear reactors, and due to the unique structure and material properties of fast reactor fuel, traditional mechanical cutting methods are not applicable. In contrast, laser cutting has emerged as an ideal alternative. However, ensuring the stability of optical fibers and laser cutting heads under high radiation doses, as well as maintaining cutting quality after irradiation, remains a significant technical challenge. Here, we study the performance changes in optical fibers exposed to a total radiation dose of 10⁵ Gy, focusing on power transmission and thermal characteristics. By integrating irradiated optical fibers with irradiated laser cutting heads, simulated cutting experiments on the hexagonal tubes of spent fuel from fast reactors (fast reactor simulation assembly) were conducted. Critical cutting quality parameters, including kerf width, surface roughness, and slagging length, were analyzed. The results indicate that, while the power transmission performance of irradiated optical fibers shows slight degradation, its impact on cutting quality is minimal. High-quality cutting can still be achieved under optimized parameters. This study confirms the feasibility of laser cutting technology in high-radiation environments and provides essential technical support for its application in nuclear fuel reprocessing. Full article

M5 cladding has emerged as a prominent fuel cladding material due to its excellent corrosion resistance. The dissolution behavior of M5 cladding is critical in both the initial cleaning stage and the reprocessing of spent fuel cladding. This study investigated the dissolution of M5 cladding in hydrofluoric–nitric (HF-HNO₃) mixed acid at varying concentrations. When the HF concentration exceeds 0.5 mol/L, the addition of strong oxidizing HNO₃ significantly reduces the dissolution rate. Moreover, HNO₃ effectively inhibits the HF-induced corrosion pitting, lowering surface roughness to 0.812 μm at a 1:5 ratio of HF:HNO₃. In addition, a surface structural analysis reveals the dissolution mechanism of M5 cladding. The β-Nb precipitated in the mixed acid was oxidized to stable Nb₂O₅ by HNO₃ while the M5 matrix surface was continuously oxidized to ZrO₂. This passivation layer inhibits further dissolution, slowing the process and enhancing the uniformity of M5 cladding. Full article

Nuclear waste is one of the most important environmental problems of nuclear power plants. A novel renewable distillation method has been proposed for the direct on-site recycling of spent nuclear fuel and the separation of its valuable components from fissile isotopes, which is especially applicable for reactors using liquid fuels. This dry separation technique can be applied in two single, parallel total-reflux columns with integrated separation stages for chlorinated nuclear waste. According to theoretical calculations, high separation accuracy of the UCl4-NpCl4, PuCl3-UCl3, CmCl3-SmCl3, and EuCl3-CsCl fractions could be achieved using twenty-six separation stages and five total-reflux repetitions, demonstrating the high efficiency of the method proposed. A scheme of the future pyroprocessing separation plant is also presented. Full article

The catalytic reaction of Tc plays a significant role in the chemical separation process during spent fuel reprocessing. However, few studies have been conducted on the chemical reaction mechanism between Tc and hydrazine. Moreover, the instability of Tc(V) and Tc(VI) makes their measurement difficult, rendering many aspects of the reaction process and mechanism unclear. This study investigates the catalytic reaction between Tc and hydrazine in a nitric acid solution. To this end, we obtained the kinetic laws of the reaction under various conditions of acidity, hydrazine concentration, and Tc concentration by monitoring the concentrations of hydrazine and Tc(VII) over time. The reaction kinetics model demonstrated that numerical simulations could effectively predict the reaction process. Results indicated that hydrazine promotes the reduction of Tc(VII) to Tc(IV), constituting the basis for establishing the Tc(IV, V, VI, VII) catalytic cycle. Among these, Tc(V) and Tc(VI) were important active intermediates and the main consumers of hydrazine. The research results may be useful for actinide separation processes based on valence control. Full article

This paper focuses on the highly radioactive, long-lasting nuclear waste produced by the currently operating fission reactors and on the sensitive issue of spent fuel reprocessing. Also included is a short description of the fission process and a detailed analysis of the more hazardous radioisotopes produced either by secondary reactions occurring in the nuclear installations or by decay of the fission fragments. The review provides an overview of the strategies presently adopted to minimize the harmfulness of the nuclear waste to be disposed, with a focus on the development and implementation of methodologies for the spent fuel treatments. The partitioning-conditioning and partitioning-transmutation options are analyzed as possible solutions to decrease the presence of long-lived highly radioactive isotopes. Also discussed are the chemical/physical approaches proposed for the recycling of the spent fuel and for the reusing of some technologically relevant isotopes in industrial and pharmaceutical areas. A brief indication is given of the opportunities offered by innovative types of reactors and/or of new fuel cycles to solve the issues presently associated with radioactive waste. Full article

This paper presents corrosion resistance results of a 12Cr ferritic ODS steel (Fe-12Cr-2W-0.5Zr-0.3Y₂O₃) fabricated via a powder metallurgy route as a prospective applicant for fuel cladding materials. In a spent nuclear fuel reprocessing facility, nitric acid serves as the primary solvent in the PUREX method. Therefore, fundamental immersion and electrochemical tests were conducted in various nitric acid solutions to evaluate corrosion degradation behavior. Additionally, polarization tests were also performed in 0.61 M of sodium chloride solutions (seawater-like atmosphere) as a more general, all-purpose procedure that produces valid comparisons for most metal alloys. For comparison, martensitic X46Cr13 steel was also examined under the same conditions. In general, the corrosion resistance of the 12Cr ODS steel was better than its martensitic counterpart despite a lower nominal chromium content. Potentiodynamic polarization plots exhibited a lower corrosion current and higher breakdown potentials in chloride solution in the case of the ODS steel. It was found that the corrosion rate during immersion tests was exceptionally high in diluted (0.1–3 M) boiling nitric acid media, followed by its sharp decrease in more concentrated solutions (>4 M). The results of the polarization plots also exhibited a shift toward more noble corrosion potential as the concentrations increased from 1 M to 4 M of HNO₃. The results on corrosion resistance were supported by LSCM and SEM observations of surface topology and corrosion products. Full article