Purex Process Research Articles

Feasibility studies on the detection of third phase formation in the solvent extraction equipment during reprocessing of fast reactor spent fuels

Fast reactor spent fuel plants reprocess plutonium rich fuel discharged from fast reactors. Centrifugal extractor or mixer settler solvent extraction equipments are employed for the selective recovery of nuclear grade uranium and plutonium from fission products. 30% (v/v) TBP/NPH is used as the solvent in PUREX process. Plutonium rich spent fuel has disadvantage of third phase formation during process upsets. Third phase can leads to flooding thereby prone for nuclear criticality or explosion hazard in evaporators. Hence, the solvent extraction equipments should be continuously monitored to detect onset of third phase. Studies are carried out by placing electrical conductivity probe in the organic outlet of a single stage mixer settler. This study measured electrical conductivity in different runs. 12 M feed nitric acid run and 15.8 M nitric acid feed acidity run indicated the conductivity at organic outlet is beyond 1035 μS/cm, indicated severe flooding as well as two phases in organic collecting container. Uranous in nitric acid run indicated when the A/O change was initiated loaded organic phase (8–11 μS/cm) conductivity at organic outlet decrease to 2.06 μS/cm as well when probe inserted at the organic and aqueous interface jumped to 190 μS/by forming Third phase. The results of this study indicated it is feasible to detect online of third phase in solvent extraction equipment such as mixer settler.

Design of experiments for the optimization of U(VI) reduction with hydrogen over Pt/SiO2

Abstract Significant amount of uranous nitrate is required for reducing Pu(IV) into inextractable Pu(III) for partitioning of U(VI) and Pu(IV) present in the loaded organic phase in PUREX process. Experiments have been conducted for the preparation of uranous nitrate by reducing uranyl nitrate present in nitric acid solution using hydrogen over Pt/SiO2 catalyst. The effect of process variables such as U(VI) concentration, H2 pressure, nitric acid concentration, catalyst quantity, temperature, mixing speed, and hydrazine concentration (for stabilizing U(IV)) on the yield of U(IV) was studied. The process variables were optimized by a two-step statistical approach namely design of experiments. The initial screening of process variables and determination of important variables that affect the production of U(IV) was determined by definitive screening design (DSD) methodology. The DSD yielded three variables affecting the U(VI) reduction to a significant extent were U(VI) concentration, H2 pressure and mixing speed. These significant variables were further optimized using five-level full factorial central composite design (CCD) methodology for understanding the intricate interactions between the variables and the combined effect of all variables at a time influencing the U(VI) reduction. A second-order polynomial equation derived from CCD was subjected to analysis of variance (ANOVA) for estimating the validity of the model and statistical significance of the terms involved in the polynomial. The results revealed that the model can predict the yield of U(IV) generation with 95% confidence in the proposed experimental range.

Effect of strontium and cesium on corrosion and contamination of 316L stainless steels during spent nuclear fuel reprocessing

The high-yield fission products of spent nuclear fuel (SNF), 90Sr and 137Cs, can deposit on the surface of facilities for SNF reprocessing plant, threatening to their safe decommissioning. To select and develop the efficient decontamination technology, the experiments for corrosion and contamination with strontium and cesium of 316 L stainless steels were performed in simulated environments of alkaline storage pond and HNO3-based PUREX process. This study analyzed the mechanism and relations of corrosion and contamination with strontium and cesium on the surface of 316 L stainless steels by combining electrochemistry and ICP-MS for the first time. When steels were exposed to the environments of alkaline and low concentration of HNO3, the presence of Sr2+ and Cs + decreased these three properties (corrosion potential, holes formed by local severe dissolution and the stability of the passive film). The results of ICP-MS demonstrated that the contamination evolution of Sr and Cs had a similar trend, and that there were significantly different contamination behaviors of steels exposed to alkaline and acidic environments. Strontium and cesium mainly deposited in the passive film of steels in both nitric acid and alkaline environments. The contamination of steels in an alkaline environment was more serious than in all nitric acid environments.

Radiation stability and radiolysis mechanism of hydroxyurea in HNO3 solution: Alpha, beta, and gamma irradiations

Hydroxyurea (HU) is a novel salt-free reductant used potentially for the separation of U/Pu in the advanced PUREX process. In this work, the radiation stability of HU were systematically investigated in solution by examining the effects of the type of rays (α, β, and γ irradiations), the absorbed dose (10–50 kGy), and the HNO3 concentration (0–3 mol L−1). The influence degree on HU radiolysis rates followed the order of the absorbed dose > the ray type > the HNO3 concentration, but the latter two had moderate effects on HU radiolysis products where NH4+ and NO2− were found to be the most abundant ones, suggesting that the differences of α, β, and γ rays should be considered in the study of irradiation effects. The radiolysis mechanism was explored using density functional theory (DFT) calculations, and it proposed the dominant radiolysis paths of HU, indicating that the radiolysis of HU was mainly a free radical reaction among ·H, eaq–, H2O, intermediates, and the radiolytic free radical fragments of HU. The results reported here provide valuable insights into the mechanistic understanding of HU radiolysis under α, β, and γ irradiations and reliable data support for the application of HU in the reprocessing of spent fuel.

Sequential Separation of Iodine Species in Nitric Acid Media for Speciation Analysis of 129I in a PUREX Process of Spent Nuclear Fuel Reprocessing.

For efficient and accurate speciation analysis of 129I in the nitric acid solution of spent nuclear fuel in its reprocessing process, a sequential procedure for stepwise separation of different iodine species in 3 mol/L HNO3 was proposed based on the solvent extraction using CCl4 and mesitylene. Molecular iodine (I2) was first separated by solvent extraction using CCl4, and iodide (I-) remaining in the aqueous phase was oxidized to I2 by adding NaNO2 and then extracted to mesitylene. Finally, iodate (IO3-) was reduced to I2 using NH2OH·HCl and extracted to mesitylene. The separation efficiency of 98-99% for tracer amounts of 129I2, 129I-, and 129IO3- in 3 mol/L HNO3 and less than 2% crossover among different iodine species were achieved. The extraction process and mechanism of different iodine species in CCl4 and mesitylene were investigated, and the problem of crossover of different iodine species due to the low extraction efficiency of low concentration of iodine in nitric acid solution was solved. A direct transfer of IO3- from HNO3 to the mesitylene phase without conversion to I2 was observed, which was attributed to the iodination of mesitylene in the HNO3 medium. Addition of a stable iodine species carrier and repeated extraction significantly improved the separation efficiency of iodine species, making their quantitative separation achievable. This method provides an approach for speciation analysis of 129I in the acidic spent nuclear fuel solution, enabling us to investigate and control the behavior of volatile 129I in the spent nuclear fuel reprocessing process.

Optimization of Process Variables for the Catalytic Reduction of U(VI) over Pt/SiO2 using Hydrazine as Reducing Agent – Design of Experiments Approach

AbstractIn PUREX process, uranous nitrate has been employed for the reduction of Pu(IV)to Pu(III).Fast generation of U(IV) with high yield is very much crucial for reprocessing of plutonium‐rich fuels.In the present work, a catalytic method for the preparation of U(IV) from U(VI) using hydrazine as a reducing agent over Pt/SiO2 catalyst surface was studied.The effect of various process variables on the efficiency of catalysis was studied and the process variables were optimized by design of experiments approach. Initial screening was carried out by Definitive Screening Design (DSD) to identify the variables that influence the reaction significantly, and later, these significant variables were optimized by Central Composite Design (CCD). A second‐order mathematical model was derived from response surface methodology (RSM) that relates the response (% reduction of U(VI)) as a function of significant variables. The model was validated experimentally to achieve 95 % U(IV) yield within 60 minutes.

Nuclear proliferation resistance assessment of fuel cycles closed with complete co-processing of spent fuel

Purex (Plutonium Uranium Recovery by Extraction) is the standard reprocessing method used for the recovery of Uranium (U) and Plutonium (Pu) in spent fuel. Products of the standard reprocessing of spent fuel discharged from a typical LWR are separated and highly pure streams of U and Pu with fissile contents of roughly 0.85 wt.% (U-235 in U) and 70 wt.% (Pu-239+Pu-241 in Pu) respectively. Because of the economic and technological conditions and safeguard concerns, alternative methods or modifications for standard reprocessing including complete co-processing have been proposed. Complete co-processing is based on the Purex process and is an easier way of recovering U and Pu from spent fuel compared to standard reprocessing. In addition, since the U and Pu recovered together from spent fuel, it does not yield pure Pu and it is advantageous with regard to safeguards. The product obtained from complete co-processing of spent fuel is a U + Pu mixture with a total fissile content of around 1.5 wt% (U-235+Pu-239+Pu-241 in U + Pu), which is not suitable for direct use in LWRs and special approaches are required for recycling it. This study focuses on the assessment of nuclear proliferation resistance of fuel cycles closed with complete co-processing and different recycling scenarios for U + Pu product. To assess the proliferation resistance, intrinsic proliferation barriers including the amount of material to be diverted for significant quantity (SQ) of Pu, isotopic content of Pu, and radiation barrier of diverted material are compared for spent fuels generated in the cycles for one GWe-yr electricity production. To assess the impact of burnup on the proliferation resistance of cycles, all calculations are performed for the burnup values 33 GWd/tHM, 40 GWd/tHM, and 50 GWd/tHM. Results showed that the closed fuel cycle in which the U + Pu product from the complete co-processing is blended with 10 wt.% enriched uranium and recycled as mixed oxide fuel possesses the highest proliferation resistance. Besides, intrinsic nuclear proliferation barriers of all the fuel cycles investigated are positively affected by an increase in burnup.

Autocatalytic decomposition of tributyl phosphate in the spent extractant of the PUREX process for safe disposal of radioactive impurities

The possibility of processing the spent extractant of the PUREX-process by thermochemical decomposition of tributyl phosphate (TBP) to hydrocarbons and phosphoric acid was investigated. It was shown that the main condition for the intensive and complete decomposition of TBP is the presence of free phosphoric acid in the system, which ensures the course of the process in an autocatalytic mode. This is observed at a molar ratio of H3PO4: TBP > 2: 1 and can be achieved in the technological process using recycled phosphoric acid. Under these conditions, at a temperature of 190–200 °C, the decomposition of TBP proceeds quantitatively within 5–10 min. It was found that the process proceeds with the formation of butene and a large amount of alkanes and alkenes of various molecular weights, which is caused by secondary processes of polymerization and cyclization of decomposition products catalyzed by excess phosphoric acid. During the decomposition of TBP, radioactive impurities pass into the phosphoric acid phase, which creates the possibility of their neutralization and disposal in the form of phosphate glasses.

Fate of Neptunium in nuclear fuel cycle streams: state-of-the art on separation strategies

Abstract Neptunium, with a half life of 2.14 million years is one of the most notorious activation products in the nuclear fuel cycle. It has been more than 5 decades in the reprocessing of nuclear fuels by the well documented PUREX process, but the fate of Np in the PUREX cycle is still not well controlled. Although Np being stable in its pentavalent state in low acid media, its starts to undergo disproportionation at higher acidities. This disproportionation along with the oxidizing conditions of the HNO3 medium makes Np to exits as Np(IV), Np(V) and Np(VI) in the dissolver solution. The overall extractability of Np in the co-decontamination step of the PUREX cycle is dependent on its oxidation state in the medium as Np(VI) and Np(IV) being extractable while Np(V) being least extractable. The present review article discusses about the speciation of Np in HNO3 and its disproportionation. The variety of redox reagents are discussed for their effectiveness towards controlling Np redox behavior in the HNO3 media. The extraction of Np with the different class of extractant has also been discussed and the results are compared for better understanding. Solid phase extraction of Np using both commercially available resin and lab based synthesized resins were discussed. The anion exchange resins with the different cationic centers were shown to behave differently towards the uptake of Np form the acidic medium. The present review also highlight the chemical conditions required for controlling or minimizing the fate of Np in different process streams of the nuclear fuel cycle.

Recovery of palladium(II) from strong nitric acid solutions relevant to high-level liquid waste of PUREX process by solvent extraction with pyrazole-pyridine-based amide ligands

The extraction of Pd2+ with N,N-dialkyl-2-(3-(pyridine-2-yl)-1H-pyrazole-1-yl) acetamide (A-C-PzPy) and N,N-dialkyl-3-(pyridine-2-yl)-1H-pyrazole-1-carboxamide (A-PzPy) from HNO3 medium, as well as the complexation of Pd2+ with A-C-PzPy were investigated. The use of tert-butyl benzene (TBB) as a diluent for A-C-PzPy offered significantly better extractability toward Pd2+, compared to that of A-PzPy. This study demonstrated strong extractability, high selectivity, high loading capacity, and fast extraction kinetics for DOct-C-PzPy toward Pd2+ from highly acidic HNO3 solution (up to 5.0 mol/L). Slope analysis showed the formation of 1:1 type of extracted species of Pd2+ with DOct-C-PzPy ligand. The extraction was a spontaneous and endothermic process with increased entropy. The analyses of electrospray ionization mass spectrometry (ESI-MS), Fourier transform infrared (FT-IR) and UV–vis spectrophotometric titration revealed that the composition of the extracted neutral complex species was Pd(NO3)2(DOct-C-PzPy). Combining the results of solvent extraction and complexation study, a neutral complexation extraction model is proposed. Furthermore, the stability constants (log β) for the complexation of Pd2+ with A-C-PzPy are also presented.

Microfluidic In-Situ Spectrophotometric Approaches to Tackle Actinides Analysis in Multiple Oxidation States.

The study and development of present and future processes for the treatment/recycling of spent nuclear fuels require many steps, from design in the laboratory to setting up on an industrial scale. In all of these steps, analysis and instrumentation are key points. For scientific reasons (small-scale studies, control of phenomena, etc.) but also with regard to minimizing costs, risks, and waste, such developments are increasingly carried out on milli- or microfluidic devices. The logic is the same for the chemical analyses associated with their follow-up and interpretation. Due to this, over the last few years, opto-microfluidic analysis devices adapted to the monitoring of different processes (dissolution, liquid-liquid extraction, precipitation, etc.) have been increasingly designed and developed. In this work, we prove that photonic lab-on-a-chip (PhLoC) technology is fully suitable for all actinides concentration monitoring along the plutonium uranium refining extraction (plutonium, uranium, reduction, extraction, or Purex) process. Several PhLoC microfluidic platforms were specifically designed and used in different nuclear research and development (R&D) laboratories, to tackle actinides analysis in multiple oxidation states even in mixtures. The detection limits reached (tens of µmol·L-1) are fully compliant with on-line process monitoring, whereas a range of analyzable concentrations of three orders of magnitude can be covered with less than 150µL of analyte. Finally, this work confirms the possibility and the potential of coupling Raman and ultraviolet-visible (UV-Vis) spectroscopies at the microfluidic scale, opening the perspective of measuring very complex mixtures.

Coprecipitation of actinide peroxide salts in the U-Th and U-Pu systems and their thermal decomposition.

The uranium and plutonium co-conversion process constitutes a continuous subject of interest for MOx fuel fabrication. Among the various routes considered, chemical coprecipitation by the salt effect has been widely investigated regarding its simplicity of integration between the partitioning and purification steps of the PUREX process, and the straightforward recovery of precursors that are easily converted into oxide phases by thermal decomposition. The present study focuses on the coprecipitation behavior of U-Th and U-Pu actinide peroxide mixed systems by examining the precipitation yields and settling properties for nitric acidity in the range of 1 to 3 M and hydrogen peroxide concentration in the range of 4.5 to 7 M. The precipitated solids have been characterized by powder XRD, IR and Raman spectroscopy, laser granulometry and SEM-EDS analyses revealing the synthesis of studtite and actinide(IV) peroxo-nitrates as aggregated particles. The actinide solid phases are uniformly distributed within the filtered cakes. The precursor thermal decomposition results in the formation of oxide phases at low temperature according to a sequential release of water molecules, peroxide ligands and nitrate ions. The calcination step has a limited effect on the morphology of the powders which remain highly divided. The high precipitation rate of actinides makes this chemical route potentially interesting as a co-precipitation process.

Involvement of 5f Orbitals in the Covalent Bonding between the Uranyl Ion and Trialkyl Phosphine Oxide: Unraveled by Oxygen K-Edge X-ray Absorption Spectroscopy and Density Functional Theory.

Monodentate organophosphorus ligands have been used for the extraction of the uranyl ion (UO22+) for over half a century and have exhibited exceptional extractability and selectivity toward the uranyl ion due to the presence of the phosphoryl group (O═P). Tributyl phosphate (TBP) is the extractant of the world-renowned PUREX process, which selectively recovers uranium from spent nuclear fuel. Trialkyl phosphine oxide (TRPO) shows extractability toward the uranyl ion that far exceeds that for other metal ions, and it has been used in the TRPO process. To date, however, the mechanism of the high affinity of the phosphoryl group for UO22+ remains elusive. We herein investigate the bonding covalency in a series of complexes of UO22+ with TRPO by oxygen K-edge X-ray absorption spectroscopy (XAS) in combination with density functional theory (DFT) calculations. Four TRPO ligands with different R substituents are examined in this work, for which both the ligands and their uranyl complexes are crystallized and investigated. The study of the electronic structure of the TRPO ligands reveals that the two TRPO molecules, irrespective of their substituents, can engage in σ- and π-type interactions with U 5f and 6d orbitals in the UO2Cl2(TRPO)2 complexes. Although both the axial (Oyl) and equatorial (Oeq) oxygen atoms in the UO2Cl2(TRPO)2 complexes contribute to the X-ray absorption, the first pre-edge feature in the O K-edge XAS with a small intensity is exclusively contributed by Oeq and is assigned to the transition from Oeq 1s orbitals to the unoccupied molecular orbitals of 1b1u + 1b2u + 1b3u symmetries resulting from the σ- and π-type mixing between U 5f and Oeq 2p orbitals. The small intensity in the experimental spectra is consistent with the small amount of Oeq 2p character in these orbitals for the four UO2Cl2(TRPO)2 complexes as obtained by Mulliken population analysis. The DFT calculations demonstrate that the U 6d orbitals are also involved in the U-TRPO bonding interactions in the UO2Cl2(TRPO)2 complexes. The covalent bonding interactions between TRPO and UO22+, especially the contributions from U 5f orbitals, while appearing to be small, are sufficiently responsible for the exceptional extractability and selectivity of monodentate organophosphorus ligands for the uranyl ion. Our results provide valuable insight into the fundamental actinide chemistry and are expected to directly guide actinide separation schemes needed for the development of advanced nuclear fuel cycle technologies.

Hydration of the pertechnetate anion. DFT study

The long-lived radionuclide 99Tc (t1/2 = 2.11 × 105 years) is one of abundant and problematic fission products in the nuclear fuel cycle with the yieldof greater than 6 % for thermal neutronfissionof uranium-235. During the reprocessing of the spent nuclear fuel (SNF) in PUREX process, 99Tc is stabilized as pertechnetate ion TcO4-. Owing to long half-life of 99Tc and high environmental mobility, TcO4- is rather environmentally hazardous radioactive pollutant. Preferentially it is to be removed at the head-stage operations when the used fuel rods are dissolved in highly concentrated nitric acid. The lack of reliable knowledge on the structure and thermodynamics of aqua complexes that the pertechnetate anion forms in aqueous solutions hampers the development of technologies for 99TcO4- separation from SNF. To fill this gap, in this paper we used DFT simulation of clusters [TcO4(H2O)n]- and {[TcO4(H2O)n-m]-[H3O]+m} (n = 1–50, m = 1–4). We have shown that unlike NO3–, SO42–, and other hydrophilic anions, the pertechnetate anion does not form a dense hydration shell in aqueous solutions. Its first hydration shell contains not more than 7 water molecules. Only three of four oxygen atoms of the anion in this shell form hydrogen bonds with water molecules. The energy of hydrogen bonds [O3TcO… H-O-H] is significantly lower (10–15 kJ mol−1) than the energy of the hydrogen bond in the water dimer (≈ 20 kJ mol−1). The hydration shell of HTcO4 preserves the same asymmetric structure in acidic solution up to acid concentration of 3 – 4 mol L–1. This suggests that during extraction in two-phase system the pertechnetate anion will predominantly concentrates at the water/organic interface, displacing more hydrophilic anions from it. When acid concentration in solution is above 4 mol L–1, water molecules are attached to the technetium atom with formation of ortho-pertechnetic acids O2Tc(OH)3 and OTc(OH)4.

SPATIAL SEPARATION OF THE IONES OF A GIVEN MASS RANGE IN THE DEMO-IMITATION SEPARATOR AT THE FIRST TURN OF IONIC TRAJECTORY

Plasma methods, where only electricity is required, are an alternative to the PUREX process used in industry for spent nuclear fuel reprocessing. It is considered the possibility of filtering out the target ions (М = 232…277) to the collector at the first turn of the ion trajectory in the plasma mass filter, which is currently being developed, that is achieved by specifying certain parameters (amplitude and frequency) for a variable component of a radial electric field. This approach significantly reduces the thermal load onto deposition surface of target ions.

High performance task-specific ionic liquid in uranium extraction endowed with negatively charged effect

Designing functionalized task-specific ionic liquid is a promising way to alleviate the potential second pollution caused by the usage of tributyl phosphate in PUREX process. In this work, a new anion-functionalized ionic liquid, N,N,N-trimethyltetradecan-1-aminium dibutyl phosphate ([N14111][DBP]), was synthesized and applied in uranium extraction. Thanks to the negatively charged effect, [N14111][DBP] exhibits high efficiency in uranium separation, at optimized condition, the maximum of distribution ratio (concentration of ions in extractant to concentration of ions in aqueous solution) for UO22+ could reach 343.5. What’s more, the selectivity of [N14111][DBP] for UO22+ from transition metal ions was also examined in simulated nuclear industry effluent containing and UO22+ ions were separated and extracted efficiently. The advantage of extraction efficiency and selectivity of [N14111][DBP] compared with DBP highlight the superiority of anion-functionalized extractant.

Dissolution behavior of fast reactor MOX nuclear fuel pellets in nitric acid medium

The fundamental information required for designing a continuous FBR MOX spent fuel dissolution system is the intrinsic kinetics of dissolution of the MOX fuel under typical PUREX process conditions. The influence of parameters such as Pu content of the MOX pellet, nitric acid concentration, intensity of mixing and reaction temperature, on the dissolution behaviour of PFBR MOX fuel pellets in nitric acid was investigated. The results clearly bring out the preferential rapid dissolution of UO 2 over PuO 2 from the same MOX pellet. This behavior was found to be inversely proportional to all the above-mentioned influencing parameters except for the Pu content of the pellets in which case it is directly related. These results will pave way for better understanding of the dissolution mechanism of MOX fuel under typical PUREX process conditions.

Electrolytic and ozone aided destruction of oxalate ions in plutonium oxalate supernatant of the PUREX process: A comparative study

The article describes an efficient and straight forward method for the destruction of oxalate ions in the plutonium oxalate supernatant stream generated during the reprocessing of FBTR spent fuel, using ozone and electrolysis without any redox intermediates/catalyst at room temperature. The influencing parameters such as concentration of nitric acid, current density and temperature were studied. The destruction was achieved in 4 and 4.5 h, respectively using ozone and electrolysis. These methods have significant advantages due to the chemically benign nature of the oxalate destructed stream for all the downstream fuel cycle processes and the feasibility to automate these process steps in a radiochemical plant.

Influence of the concentration of nitric acid on the composition of NOX gas evolved during the dissolution of nuclear fuel and its implications on the PUREX process

Understanding the mechanism of dissolution of UO2 in nitric acid is important for designing a continuous dissolution system for the reaction. Knowledge of the composition of the NOX gas evolved during the course of the reaction is imperative for this understanding as it strongly influences the reaction mechanism. In this work, the profiles of the composition of NOX gas evolved and the ratio of moles of acid consumed per mole of uranium dissolved during the dissolution of UO2 pellets in nitric acid have been determined as a function of acidity. The predominant stoichiometric reaction by which UO2 dissolves at every point of the reaction was clearly identified thereby improving the understanding of its mechanism and facilitates better control over the overall reprocessing process flowsheet.

Prospects of functionalized porous materials towards extraction of spent nuclear fuel radionuclides

Radioactive waste levels have continued to increase due to the growth and development of nuclear energy, industrial or medical radioactive use. In this regard, conventional radioactive waste generated by nuclear power plants cannot be ignored. Effective management of nuclear radioactive waste plays a vital role in alleviating negative impacts on the society and environment. Despite the progress that has been made concerning separation and recycling of spent nuclear fuel through the PUREX process, several gaps in knowledge still exist especially towards the development of a robust separation system based on solid-phase extraction using porous materials. Solid phase extraction is being viewed as one of the most convenient and effective approaches in the removal of cations in radionuclide solutions. This is due to its ability to increase the selectivity and sensitivity of the method as it permits discriminatory binding of analyte to a solid support where the analyte can be collected and thereafter eluted using small quantity of a different solvent. The review covers the current methods used in aqueous nuclear reprocessing, highlights their deficiencies and introduces the potential of applying solid-phase extraction in management of nuclear waste. This study gives the prospects of functionalized porous sorbent materials as solid support in solid-phase extraction of spent nuclear fuel elements.