Purex Process Research Articles

Partitioning of the high-level radioactive waste from the purex process by 40% TBF in isoparaffin using magnesium nitrate as a salting-out agent

Centrifugal contactor rig trials of the partitioning flowsheet were carried out using simulated and real high level waste solutions obtained by reprocessing of WWER-100 spent nuclear fuel with burn-up of 47 MW day/t. 40% TBP in Isopar M was used as a solvent with magnesium nitrate as a salting-out agent. Complete removal of Cs, Sr, Zr and Mo to raffinate was achieved. Rare earth elements, Am and Cm were totally concentrated in TPE and REE product. Comparison of the results of this work with previous trials with iron (III) nitrate as a salting-out agent [1] demonstrated that using magnesium nitrate as a salting-out agent resulted in higher decontamination factors of TPE and REE fraction from Cs (~10000) than using iron (III) nitrate (7500).

Comparison of dynamic tests of the modified purex process flowsheet using 30 and 45% TBF in isoparafin

Results of centrifugal contactor rig trials of the flowsheet of the first solvent extraction cycle in protective glove boxes using simulated solutions of spent nuclear fuel and 30 and 45% TBP in isoparaffin as a solvent are compared. It was shown that an increase in the TBP concentration to 45% resulted in deterioration of the quality of the products obtained by the same flowsheets with 30% TBP in Isopar L as a solvent. Comparison of experimental and computer simulated profiles of component distribution across the stages of extraction units demonstrated satisfactory agreement in nitric acid and uranium concentrations. At the same time, it was found that simulation model of zirconium, technetium, and neptunium extraction requires improvement considering interaction of these elements with complexing agents.

Orthogonal signal correction assisted multivariate regression approach for the estimation of uranium and acidity in PUREX process streams

A direct UV–Visible absorbance spectrophotometric method was developed for the simultaneous determination of uranium and nitric acid concentration in the PUREX process samples. The simulated system consisted of uranium and nitric acid in concentration range corresponding to reprocessing of spent nuclear fuel discharged from nuclear reactor was prepared. The absorbance of these samples was measured in the range of 400–470 nm at a scan speed of 100 nm/s and resultant spectra were recorded. The changes in wavelength maxima of U(VI) absorption spectrum at different nitric acid concentration was utilized to determine the concentration of uranium and nitric acid in the sample by orthogonal signal correction assisted principal component regression. After the principle component regression the RMSEP for test data (Uranium: 3–21 g/L and acidity: 2–12 M) were 0.7 g/L and 0.4 M respectively. This method is superior to conventional method being followed for routine analysis of plant control samples in view of minimizing the generation of radioactive analytical waste consisting other corrosive reagents and reducing radiation exposure to operators during analysis. This method is amenable for online monitoring also.

Densities and Molar Volumes for Binary Solutions of Tri-iso-Amyl Phosphate and n-Dodecane in the Temperature Range of 283.15–363.15 K

Tri-n-butyl phosphate (TBP) has remained workhorse of PUREX based nuclear recycle operations since 1956. However due to its associated disadvantages like third phase formation and tendency to react violently with nitric acid at elevated temperatures generated requirement of an alternate extractant for PUREX process. Tri-iso-amyl phosphate (TiAP) has better attributes than TBP on both the counts. Density of solutions is an important process parameter for design of equipment, process operations and control and safety analysis. In this work, experimental data on the density of binary solution of tri-iso amyl phosphate and n-dodecane are reported for several temperature levels. The related volumetric parameters were also estimated from experimental data as well as Kirkwood–Buff integrals. Derivative parameters like linear coefficients of preferential solvation δij(i≠j)o\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$\\delta_{ij(i \ e j)}^{o}$$\\end{document} and local mole fractions xij(i≠j)\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$x_{ij(i \ e j)}$$\\end{document} at 298.15 K were also estimated and reported.Graphical

Synthesis, Characterization and Dissolution behavior of Nitrides of U-Zr and U-Pu-Zr metallic alloy fuels for Aqueous reprocessing

The aqueous based PUREX process is an alternate choice to pyro-reprocessing of metallic alloy fuels (U-Zr/U-Pu-Zr). Direct dissolution of metallic alloy fuels in a nitric acid medium is not feasible due to vigorous reaction exhibited by the alloys when exposed to this medium. Conversion of metallic alloys into their oxides is a challenging method due to poor dissolution of oxides in nitric acid. Excellent dissolution of nitrides compared to oxides of metallic alloy fuels lead to exploit the nitride route for aqueous reprocessing. Hence, a novel method was adopted to prepare the nitrides of unirradiated U-Zr and U-Pu-Zr alloys via hydridation using UH3, as source-cum-getter instead of hydrogen cylinder. In this context, nitrides of unirradiated U-Zr/U-Pu-Zr metallic alloys were prepared by nitridation via hydridation. Initially, nitrides of U and Zr metals were synthesized and the formation of ZrN and U2N3 and UN phases were confirmed by powder X-ray diffraction measurements. Subsequently, nitrides of unirradiated U-Zr and U-Pu-Zr alloys were synthesized and characterized by X-ray diffraction, scanning electron microscopy and particle analyzer. The dissolution behavior of nitrides of Zr, U, U-Zr and U-Pu-Zr alloys were examined in nitric acid (12 M) medium under reflux conditions at 383 K for 8 h. The present study highlights the potential of the nitride route for efficient dissolution of actinides (U and Pu) and it is very important to develop flow sheets for the aqueous reprocessing of metallic alloys fuels.

Partitioning of the High Level Radioactive Waste from the Purex Process by 40% TBF in Isoparaffin Using Magnesium Nitrate as a Salting-out Agent

Partitioning of the High Level Radioactive Waste from the Purex Process by 40% TBF in Isoparaffin Using Magnesium Nitrate as a Salting-out Agent

Comparison of Dynamic Tests of the Modified Purex Process Flowsheet Using 30 and 45% TBF in Isoparaffin

Comparison of Dynamic Tests of the Modified Purex Process Flowsheet Using 30 and 45% TBF in Isoparaffin

Determination of concentration of formaldehyde in high level waste during the destruction of nitric acid in PUREX process

Determination of concentration of formaldehyde in high level waste during the destruction of nitric acid in PUREX process

Nuclear Waste Management

Nuclear energy production generates nuclear waste. Nuclear and radioactive waste, especially high level waste (HLW) and intermediate level waste (ILW), require special long-term safe management solutions. One part of the solution involves recycling through reprocessing of spent nuclear fuel. Recycling reduces the volume of nuclear waste and allows for the reuse of some of its components. This can be achieved through methods such as adsorption, ion exchange, coagulation, flotation, filtration, chemical precipitation, reverse osmosis, and solvent extraction, like the PUREX process. Another possible future solution involves partitioning and transmutation technology, which can reduce the production of nuclear waste. The best long-term solution is the immobilisation of HLW and ILW in a solid matrix. Materials used for this purpose include glasses, cements, bitumen, geopolymers, concrete, and ceramics as a promising material. While cementation is still the most commonly used immobilisation method due to its low cost and simplicity, vitrification is a more permanent long-term solution. Deep geological disposal in combination with vitrification and a robust multi-barrier system is considered the most acceptable solution for safe nuclear waste isolation. This review provides insight into the mostly commonly used and promising immobilisation materials, as well as the most effective methods and technologies currently in use and under development for the management of HLW and ILW in nuclear waste management.

Using life cycle assessment to quantify the environmental benefits of circular economy strategies in the nuclear industry

In the United Kingdom nuclear energy is expected to play a key role in decarbonising the power generation sector. The implementation of circular economy approaches in the nuclear industry can minimize the amount and the associated environmental impacts of nuclear wastes. In this article, we demonstrate how Life Cycle Assessment (LCA) can be used to investigate the environmental benefits of two circular economy strategies that aim at reducing the amount of intermediate level waste (ILW) destined for disposal in a geological disposal facility. The first case study focuses on a novel technology for recycling zirconium alloy cladding waste, whilst the second case study investigates the environmental benefits of using depleted (instead of natural) uranium to produce uranyl nitrate, a chemical used in the PUREX process. Our results show that both circular approaches outperform conventional ones across all environmental categories and particularly in terms of resources depletion, with reductions up to 25% and 94% for respectively zirconium alloy recycling and depleted uranium reuse. The environmental benefits of both approaches are due not only to a reduction in the amount of ILW to be disposed of, but also because they are assumed to induce a reduction in the demand for mining of primary zirconium or uranium respectively. When both approaches are combined, the environmental benefits range from 4% in the category freshwater and up to 94% in the category resource use, energy carriers.

The effect of size and shape of annular centrifugal contactors upon hydrodynamics and mass transfer using HNO3-TBP/OK

Annular centrifugal contactors (ACCs) are process intensification equipment for liquid-liquid extraction (also known as solvent extraction) that use centrifugal force that 200x greater than traditional gravity settling (Baker et al., 2022a). The size and throughput of ACCs have been denoted by their increasing rotor diameter (RD), from 10 mm up to 1330 mm. In this work a multiscale ACC platform has been commissioned, (consisting of 10, 25, 40 mm RD ACCs), to collect the physical liquid-holdup and mass transfer across several sizes to inform future computation modelling flowsheets of liquid-liquid extraction (LLE).Experimental studies have been performed with nitric acid with tributyl phosphate in odourless kerosene (TBP/OK) relevant to the development of advanced PUREX processes for recovery of actinides from spent fuel. Hydrodynamic trials determined the minimum Relative Centrifugal Force (RCF) required to separate extreme solvent/aqueous (S/A) feed ratios and total throughputs across the three sizes of ACC. The novel 'stepped stop' method has provided understanding of the actual liquid holdups within the two distinctly different regions (mixing annulus and separating rotor) ACC during operation to provide accurate residence times. These values were subsequently used to calculate the mass transfer and stage efficiency of aqueous nitric acid with tributyl phosphate in odourless kerosene. The smallest contactor was determined to be distinct and likely under-predicting performance of larger contactors. Ultimately the geometry and the aspect ratio supersedes using rotor diameter for defining scale-up of annular centrifugal contactors.

An Integrated Model Combining Mass Transfer and Chemical Reaction for Co-Decontamination Extraction Step of PUREX Process in a Pulsed Extraction Column

ABSTRACT The accurate prediction of the extraction behaviors of various solutes in PUREX reprocessing process is crucial for the operation and implementation of the actual process. In this paper, an integrated model is developed to predict the extraction behaviors of U, Np, Pu, and HNO3 in the co-decontamination step (1A extraction step) in a pulsed extraction column. The model couples several physical and chemical processes, such as countercurrent flow, mass transfer, and chemical reaction. The mass transfer coefficients and the distribution ratios of U(VI), Pu(IV), Np(IV), Np(V), Np(VI), HNO3 and HNO2 can be obtained using this model. In particular, the redox and disproportion reactions of Np are considered in the model, and the flow direction of Np can be judged under various process conditions and the corresponding influential factors can be analyzed. The judgment is based on the yield calculated from the relative concentration profiles of Np(VI), Np(V), and Np(IV) in the two phases. For neptunium to inter 1AP step, it is necessary to select high nitric acid concentration, low nitrite concentration and especially high flow ratio. For neptunium to inter 1AW step, low nitric acid concentration, high nitrite concentration and low flow ratio are needed. Compared with the experimental data, the relative errors of the distribution ratios of various solutes are less than 30%, and the relative errors of the concentrations of U(VI) and Pu(IV) at the outlet of the organic phase are less than 10%, and our model is indicated to be reliable and applicable for co-decontamination step in the PUREX reprocessing process.

The rapid photochemical reduction of U(VI) at high uranium concentrations relevant to spent nuclear fuel recycle processes

An Advanced PUREX process for the recycling of spent nuclear fuel is currently under active development in the UK. Its key aims are to avoid pure separated plutonium at all stages of the process to enhance the level of proliferation resistance, and to achieve a single cycle flowsheet that has a smaller plant footprint with consequent decreases in the capital cost and secondary wastes generated. Addressing these aims, a significant feature of the process is the co-treatment of U and Pu and thus the in situ co-conversion of mixed actinide metal nitrate solutions into oxide powders, suitable for the fabrication of new mixed metal oxide (MOx) fuel. The baseline industrial process for plutonium recovery is by oxalate precipitation; however, in order to quantitatively recover both U and Pu the uranium must be in the U(IV) oxidation state due to the high solubility of U(VI) oxalate. The first stage of this co-conversion, the development of which is reported on here, is the rapid, clean photochemical co-reduction of a mixed U(VI)/Pu(IV) nitrate stream to U(IV)/Pu(III).Here we describe a study of the reduction of U(VI) in preparation for mixed U(VI)/Pu(IV) reduction trials. Exploiting the photochemistry of U, we demonstrate the convenient and efficient photo-excitation and chemical reduction of U(VI) upon exposure to 407 nm wavelength light in the presence of alcohol-based reductants. Using a purpose built laboratory-scale photochemical reactor, U(VI) solutions of up to process-relevant concentrations of 630 mmol/dm3 (150 g/l) U have been successfully converted to a U(IV) product, achieving a conversion efficiency of ∼98% within 1.5–15 min when using propan-2-ol as a sacrificial reductant. Modelling of the dependence of the rate of U(IV) generation on initial U(VI) concentration reveals the importance of light penetration depth and effective solution mixing in determining the efficiency of the photochemical process at high U-loadings. It also reveals that the photoreduction of U(VI) to U(IV) occurs by two sequential 1-electron reductions: (i) the photochemically driven reduction of UO22+ to UO2+ by propan-2-ol, which itself is oxidised to form an α-hydroxyalkyl radical, immediately followed by (ii) a second chemical reduction of UO22+ to UO2+ and/or UO2+ to U4+ by the so-formed radical. With the addition of a nitrous acid scavenger to prevent re-oxidation of the photochemically generated U(IV), a stable product is maintained indefinitely, and the solution is suitable for subsequent oxalate co-precipitation as part of a MOx fuel fabrication process.

Irradiation of Plutonium Tributyl Phosphate Complexes by Ionizing Alpha Particles: A Computational Study.

The PUREX solvent extraction process, widely used for recovering uranium and plutonium from spent nuclear fuel, utilizes an organic solvent composed of tributyl phosphate (TBP). The emission of ionizing particles such as alpha particles, resulting from the decay of plutonium, makes the organic solvent vulnerable to degradation. Here, we study the ultrashort time alpha irradiation of tributylphosphate (TBP) and Pu(NO3)4(TBP)2 complex formed in the PUREX process. Electron dynamics is propagated by Real-Time-Dependent Auxiliary Density Functional Theory (RT-TD-ADFT). We investigate the use of previously proposed absorption boundary conditions (ABC) in the molecular orbital space to treat secondary electron emission. Basis set and exchange correlation functional effects with ABC are reported as well as a detailed analysis of the ABC parametrization. Preliminary results on the water molecule and then on TBP show that the phenomenological nature of the ABC parameters necessitates selecting appropriate values for each system under study. Irradiation of free and complexed TBP shows an influence of the ligands on the variation of atomic charges on the femtosecond time scale. An accumulation of atomic charges in the alkyl chains of TBP is observed in the case where the nitrate groups are predominantly irradiated. In addition, we find that the Pu atom regains its electric charge very rapidly after being hit by the projectile, with the coordination sphere serving as an electron reservoir to preserve its formal redox state. This study paves the road toward a full understanding of the degradation of organic extracants employed in the nuclear industry.

Extractive Recovery and Сoncentration of Palladium(II) from Simulated Nitrate–Nitrite PUREX Process Raffinate with 4-[(Hexylsulfanyl)methyl]-3,5-dimethyl-1-phenyl-1H-pyrazole

Extractive Recovery and Сoncentration of Palladium(II) from Simulated Nitrate–Nitrite PUREX Process Raffinate with 4-[(Hexylsulfanyl)methyl]-3,5-dimethyl-1-phenyl-1H-pyrazole

Removal Effectiveness for Washing Ce(IV), U(VI), and Zr(IV) Ions from Contaminated Purex Solvent by Hydrazine Carbonate

The contaminated solvent from the Purex process is washed with alkaline detergents such as sodium carbonate, which generates a large amount of secondary wastes. Therefore, hydrazine carbonate as a salt-free reagent deserves to be studied in depth. In this study, the Ce(IV), U(VI), and Zr(IV) metal ions in organic phases containing dibutyl phosphate (HDBP) of 30% tributyl phosphate (TBP)–dodecane were washed with hydrazine carbonate. The effects of the oscillation time (1 to 15 min); temperature (25°C to 85°C); cumulative number of washes (one to four times); mass fraction of hydrazine carbonate (0.1% to 20%); volume ratio of the aqueous phase to the organic phase (0.2 to 5); HDBP concentration (0 to 0.4 M); HNO3 concentration (0.05 to 8 M); and concentration of Ce(IV), U(VI), and Zr(IV) metal ions on the removal percentages of Ce(IV), U(VI), and Zr(IV) metal ions in polluted solvents were studied. The results showed that when the organic phase containing 0.02 M HDBP was washed three times with 5% hydrazine carbonate at 25°C, the removal percentages of the Ce(IV), U(VI), and Zr(IV) ions were 96%, 98%, and 94%, respectively. Meanwhile, the retention concentrations of the three in the organic phase were 35, 28, and 78 mg/L, respectively. The increase of the mass fraction of hydrazine carbonate enhances the removal of the metal ions from the organic phase into the aqueous phase. High acid is not conducive to alkaline washing of metal ions. The increase of HDBP concentration not only promotes extraction but also increases the retention capacity of the organic phase and has the most significant effect on Zr(IV). U(VI) promotes the preferential washing of Zr(IV) while Ce(IV) increases the metal retention concentration of Zr(IV) in the organic phase.

Beyond U/Pu separation: Separation of americium from the highly active PUREX raffinate

This review provides a comprehensive overview of published hydrometallurgical chemical processes capable of separating americium from curium and the lanthanides. A search for highly selective and robust americium separation methods is motivated by the fact that americium isotopes contribute significantly to the long-term heat load and residual radiotoxicity of high level waste originating from the PUREX process, nowadays still the key reprocessing technology to recycle uranium and plutonium from spent nuclear fuel. The separation (partitioning) and subsequent nuclear transmutation (or burning) of americium would allow a substantial improvement in the construction of an underground final repository and provide safety benefits – highly relevant for the exploitation of such a facility over extended periods of time. Besides the discussion of basic properties of the various separation methods, an evaluation of their compatibility with upstream and downstream processes as well as the treatment of secondary waste streams is also provided.

Substantiation of the Practical Feasibility of the Modified Scheme of Purex Process with an Increased TBP Concentration Using Mathematical Modeling

Substantiation of the Practical Feasibility of the Modified Scheme of Purex Process with an Increased TBP Concentration Using Mathematical Modeling

Study on the benefit of HLLW partitioning on the high-level waste glasses from the viewpoint of waste management

The volume and characteristics of heat-generating waste glasses produced through spent fuel reprocessing PUREX process + HLLW partitioning TRPO process are evaluated as a function of burn-up and cooling period of spent fuel in this study. The waste loading of glass is assumed to be restricted by the heat generation rate, MoO3 content, and noble metal content. The results show that: 1) For spent fuel of 8–20 years cooling, the volume of waste glasses after HLLW partitioning is 39%–25% of that before partitioning. 2) After removing >99.967% of α-nuclides in HLLW through TRPO process, the TRPO raffinate stream can be immobilized into waste glasses. After 95 years of decay storage, intermediate-depth disposal rather than geological disposal of TRPO raffinate waste glasses could be engineered. 3) Separation of Sr and Cs from TRPO raffinate hardly reduces the volume of waste glasses. SrCs separation before vitrification is unnecessary. 4) The insoluble residues generated during the shearing and dissolution of spent fuel may not be suitable for mixing with the TRPO raffinate stream, nor for immobilizing into a borosilicate glass matrix. Alternative solutions rather than vitrification are encouraged.

Development of a three-stage, cross-current air sparged mixer-settler for the purification of degraded PUREX solvent

ABSTRACT Tri-butyl phosphate (TBP) diluted with in n-dodecane (n-DD) is used as the solvent in PUREX process. Solvent degradation due to radiolysis and hydrolysis during the process is one of the major problems prevailing in reprocessing plants. This paper presents an overview of the design, fabrication, and performance evaluation of a three-stage, continuous, cross-current air sparged mixer-settler developed for purifying the degraded PUREX solvent using hydrazine carbonate. The developmental program consisted of various stages, starting with small-scale batch experiments to investigate the solvent treatment kinetics and phase separation characteristics, followed by design fabrication and performance evaluation of three-stage air sparged mixer-settler. The study on solvent treatment kinetics in a 1.6 L batch air sparged mixer using 1.5 M hydrazine carbonate showed that the concentration of di-butyl phosphate (DBP) in the simulated degraded solvent could be reduced from 8,000 mg L−1 to less than 40 mg L−1 within 180 seconds. Performance evaluation of the three-stage air sparged mixer settler had shown that the DBP concentration in the simulated degraded organic phase could be brought down from 8,000 mg L−1 to less than 15 mg L−1, and the equipment could process up to 60 L h−1 throughput with negligible phase entrainment.