Ion‐Exchange Separation of Pu from Macro Concentration of Th in HNO3 Medium

Abstract

Sorption behavior of Th and Pu from anion‐ as well as cation‐exchange resin was investigated from nitric acid medium by both batch and column methods. The anion‐exchange studies involved anionic nitrate complexes of Pu⁴⁺ and Th⁴⁺ sorbed onto DOWEX 1x4 resin (50–100 mesh), and the cation‐exchange studies involved the sorption of Pu³⁺ and Th⁴⁺ onto BIORAD AG 50Wx8 (50–100 mesh) or DOWEX 50Wx4 (50–100 mesh) resin. The batch data gave a separation factor (K _d,Pu/K _d,Th) of 22 for the anion‐exchange method and 0.017 for the cation‐exchange method at 3 and 2 M HNO₃, respectively. A two‐stage ion‐exchange separation method was developed for the quantitative separation of Pu (8 g/L) from a macro amount of Th (200 g/L) in nitric acid medium. The first step involved the quantitative sorption of plutonium from the mixture while about 90% of Th could be washed in 6 column volumes. The plutonium, eluted (as Pu³⁺) using 0.5 M HNO₃ + 0.2 M hydrazinium nitrate (HN) + 0.2 M hydroxyl ammonium nitrate (HAN), and the residual (～10%) Th were subsequently loaded onto a cation‐exchange column in the second step. Greater than 99% Pu was recovered with 2 M HNO₃ (in ～8 column volumes) containing 0.2 M HN + 0.2 M HAN. The final elution of thorium from the cation‐exchange column was achieved in about 6 column volumes of 1 M α‐hydroxy isobutyric acid. A (Pu, Th)O₂ fuel scrap sample was dissolved in 16 M HNO₃ containing 0.005 M HF and was used subsequently as the feed for the anion‐exchange column. The eluted Pu was subsequently loaded onto a cation‐exchange column for final purification. The recovery of plutonium and thorium was found to be >99% and >98%, respectively, while the respective decontamination factors were estimated to be 215 and 292.     

Solvent Extraction of Plutonium(IV), Uranium(VI), and Some Fission Products with Di-n-octylsulfoxide

Abstract

Extraction behavior of plutonium(IV), uranium(VI), and some fission products from aqueous nitric acid media with di-n-octylsulfoxide (DOSO) has been studied over a wide range of conditions. Both the actinides are extracted essentially completely, whereas fission product contaminants like Zr, Ru, Ce, Eu, and Sr show negligible extraction. The absorption spectra of sulfoxide extracts containing either Pu⁴⁺ or UO₂ ²⁺ indicate the species extracted from nitric acid into the organic phase to be Pu(NO₃)₄. 2DOSO and UO₂(NO₃)₂. 2DOSO, respectively. Extraction of these actinides decreases with increasing temperature, indicating the extraction to be exothermic. DOSO extracts plutonium and uranium better than di-n-hexylsulfoxide (DHSO) under all condition and is also more soluble in aromatic diluents than the latter. The effect of gamma radiation on the extraction properties of DOSO is found to be similar to that of DHSO.     

Application of N,N-Dialkyl Aliphatic amides in the Separation of Some Actinides

Abstract

N, N-dialkyl substituted alkyl amides are known to be good extractants of some actinides such as U, Pu, and Th. Their stability is comparable to that of TBP, and their degradation products do not interfere as do the degradation products of TBP. On the other hand, the principal disadvantage of the amides is their tendency to form poorly soluble U adducts in organic diluents.

A systematic investigation has been carried out on the extractive behavior of two typical alkyl amides of different structures with respect to the actinide ions UO₂ ²⁺, Th⁴⁺, Np⁺⁴, Pu⁺⁴, NpO₄₊ ², PuO²⁺ ₂, Pu³⁺, and Am³⁺, as well as with respect to the most significant fission products. The results obtained have been compared with those obtained using TBP in the same experimental conditions, verifying the applicability of amides in the separation of U from Th.     

Selective Facilitated Transport of Uranium(VI) Across a Bulk Liquid Membrane Containing Benzoyltrifluoroacetone as Extractant-Carrier

Abstract

A study has been made on carrier-mediated transport of uranium(VI) using a bulk liquid membrane prepared by dissolving benzoyltrifluoroacetone (HBTA) in carbon tetrachloride. The source phase comprised of a solution of UO₂ ²⁺ or its binary mixture with other cations such as Th⁴⁺, Hf⁴⁺, Zr⁴⁺, Fe³⁺, La³⁺, Cu²⁺, Co²⁺, Mn²⁺, Ni²⁺, and Zn²⁺ in aqueous solutions of pH 6.0, while 0.1 M hydrochloric acid was serving as a stripping agent in the receiving compartment. The interference from Th⁴⁺ and a few other cations could be eliminated by using trans-1,2-diaminocyclohexane-N,N,N′,N′-tetraacetic acid (DCTA) as a proper masking agent in the feed solution. Various factors influencing the transport process have been studied and an uphill transport (>99%) of uranium(VI) from the source phase could be accomplished under optimum conditions.     

Influence of the PuO2 content on the sintering behaviour of UO2-PuO2 freeze-granulated powders under reducing conditions

The sintering behaviour of freeze-granulated UO₂-PuO₂ powders containing 33 and 15 mol% Pu/(U + Pu) was investigated under reducing conditions up to 1700 °C. For both compositions, the “grain size versus relative density” trajectory was constructed. All the experimental points form a single trajectory meaning that a relative density/grain size pair obtained after sintering seems independent of the thermal path (heating rate, soak time, soak temperature) and of the Pu content. Exploiting the “grain size versus relative density trajectory” enabled also to propose that densification was controlled by grain boundary diffusion and grain growth by the grain boundaries whatever the Pu content. An activation energy around 510 kJ/mol was obtained for densification, which was close to the value reported for the grain boundary diffusion of plutonium cations in U_1-xPu_xO₂ polycrystals. Whatever the Pu/(U + Pu) content, the sintered microstructure of 98 % dense samples possesses a homogeneous distribution of plutonium and uranium cations.     

Extraction of Uranium and Thorium from Nitric Acid Solution by TODGA in Ionic Liquids

Extraction of uranium (UO₂²⁺) and thorium (Th⁴⁺) from a nitric acid solution into an imidazolium-type ionic liquids (ILs) of 1-alkyl-3-methylimidazolium hexafluorophosphate ([C_nmim][PF₆], n = 6 or 8) was carried out using N,N,N′,N′-tetraoctyl-3-oxapentanediamide (TODGA) as an extractant. It was found that the extraction efficiencies of UO₂²⁺ and Th⁴⁺ ions are higher in comparison with that done in n-dodecane. The extraction mechanism was deduced by the slope analysis and extraction experiment. Transfer of both ions is assumed to proceed predominantly through the neutral solvation mechanism from nitric acid solution into ILs. The UO₂²⁺ ion forms a 1:2 complex with TODGA in ILs at lower acidity, and a 1:1 complex in ILs and in n-dodecane at higher acidity. The Th⁴⁺ ion forms a 1:2 complex with TODGA in C₆mimPF₆ IL or a 1:1 complex in C₈mimPF₆ IL at lower acidity and a 1:1 complex in both ILs, and n-dodecane at higher acidity. Stripping studies were conducted using sodium salt of EDTA as a stripping ligand. The thermodynamics of extracting UO₂²⁺ ions and Th⁴⁺ ions from a 3 M HNO₃ solution was also studied. The results indicated that the extraction reactions are spontaneous and go through an exothermic process.     

Aqueous Biphase Systems for Liquid/Liquid Extraction of f-Elements Utilizing Polyethylene Glycols

Abstract

Aqueous biphasic systems formed by adding a H₂O soluble polymer (polyethylene glycol) to an aqueous salt solution ((NH₄)₂SO₄ or K₂CO₃) have been investigated for use in extracting aqueous Am³⁺, Pu⁴⁺, UO₂ ²⁺, and Th⁴⁺ ions into the polymer-rich phase. Extraction occurs only in the presence of complexing dyes which preferentially partition to the polymer-rich phase. Three such dyes, arsenazo III, alizarin complexone, and xylenol orange were investigated. Arsenazo III extracts all four metal ions from SO₄ ^2- media but not from CO₃ ^2- solutions. Alizarin complexone quantitatively extracts Th⁴⁺ and Pu⁴⁺ from SO₄ ^2- media, while Am³⁺ is the best extracted ion in CO₃ ^2- solution. Xylenol orange extracts only Am³⁺ from CO₃ ^2- media. In SO₄ ^2- solutions low concentrations of xylenol orange extract Th⁴⁺ and Pu⁴⁺, while Am³⁺ and UO₂ ²⁺ are extracted at higher concentrations of xylenol orange. H₂SO₄ can be used to strip the metal ions, while NH₄OH often but not always enhances the extraction.     

Removal of the tetravalent actinide thorium from solution by a biocatalytic system

The aim of this work was to study the effects of pH, a metal complexing reagent (citrate) and the concentration of substrate on the bioaccumulation of tetravalent actinides, using Th⁴⁺ as a model, by a phosphates-catalysed reaction. This yields HPO which precipitates with heavy metals as cell-bound metal phosphate. Poor removal of The from solution was observed, which is in accordance with the solution chemistry of the metal. A considerable improvement in the efficiency of thorium removal was obtained by incorporating ammonium acetate (NH₄Ac) into the solution. Although extensive deposition of polycrustaling NH₄UO₂PO₄ was observed previously by cells that had accumulated UO, no evidence for deposition of cstalline thorium phosphate was obtained by X-ray diffraction analysis. Examination by proton induced X-ray emission (PIXE) analysis. Examination by proton induced X-ray emission (PIXE) analysis showed a non-homogenous thorium deposit of variable phosphorous content.     

Recent R&D Studies Related to Coprocessing of Spent Nuclear Fuel Using N,N-Dihexyloctanamide

Abstract

The PUREX process has undergone several modifications to address the issues of high burn up, fewer solvent extraction cycles, and reduced waste arisings. Advanced fuel cycle scenarios have led to a renewed international interest in the development of separation schemes for co-recovering U/Pu from spent fuels. Completely incinerable N,N-dihexyloctanamide (DHOA) has been identified as a promising candidate for the reprocessing of spent fuels. Batch extraction studies were carried out to evaluate DHOA and TBP for the coprocessing (co-extraction and co-stripping) of U and Pu from spent fuel under varying concentrations of nitric acid and of uranium as well as under simulated pressurized heavy water reactor spent fuel feed conditions. At 50 g/L U in 4 M HNO₃, D_Pu values for 1.1 M DHOA and 1.1 M TBP solutions in n-dodecane were 7.9 and 3.8, respectively. In contrast, significantly lower D_Pu value at 0.5 M HNO₃ (4 × 10^?3) for DHOA as compared to TBP (4 × 10^?2) suggested that it was a better choice for coprocessing of spent nuclear fuel. This behavior was attributed to the change in stoichiometry of extracted species at lower acidity vis-a-vis the higher acidity. These studies suggest that plutonium fraction can be enriched with respect to uranium contamination in the product stream. DHOA displays better extraction behavior of plutonium and stripping behavior of uranium under simulated feed conditions. DHOA appears distinctly better than TBP with respect to fission product/structural material decontamination of U/Pu.     

Extraction of HNO3 and some ions with tributyl phosphate (tbp)-mixed trialkyl phosphine oxides (trpo)/kerosene

Tributyl phosphate (TBP) and trialkyl phosphine oxides (TRPO) are important extractants. They are widely used in industrial extraction processes, especially in the nuclear power industry. However, both TBP and TRPO suffer from several disadvantages. TBP has a low extractability for trivalent transuranium elements such as Am³⁺ and Pu³⁺ while TRPO has low loading capacity for HNO₃ and UO₂ ²⁺. The extraction of HNO₃ and 20 other ions of importance in the nuclear power industry was studied using TBP-TRPO/kerosene. The loading capacity of UO₂ ²⁺ and HNO₃ in TBP-TRPO/kerosene was determined. The synergistic extraction characteristics of the mixture for Am³⁺ and TcO₄ m were studied. The influence of high-concentration UO₂ ²⁺ on the extraction of Am³⁺, Eu³⁺, Pu⁴⁺, and TcO₄ m was investigated. The experimental results show that TBP-TRPO/kerosene mixtures display both a high extractability for a number of ions and a high loading capacity for UO₂ ²⁺ and HNO₃.     

Highly efficient extraction of tetra- and hexavalent plutonium using DGA functionalized pillar[5]arene in RTIL: Understanding speciation,thermodynamics and radiolytic stability

DGA functionalized pillar[5]arene (P5DGA) in ionic liquid was demonstrated as highly efficient system for the extraction of plutonium from acidic aqueous solution in tetravalent and hexavalent oxidation state. The extraction followed ‘cation-exchange’ mechanism via [Pu.P5DGA]⁴⁺ and [PuO₂.P5DGA]²⁺, as extracted species for Pu⁴⁺ and PuO₂²⁺, respectively. Evaluation of thermodynamic parameters (ΔG, ΔH and ΔS) showed the feasibility and spontaneity of the extraction process. The process was exothermic and primarily ‘enthalpy driven’, since entropy change was found negative. P5DGA-RTIL solvent system showed good radiolytic stability even at 1000 kGy of gamma dose.     

Silica nanoparticles modified with a Schiff base ligand: An efficient adsorbent for Th(IV), U(VI) and Eu(III) ions

Modification of SiO₂ nanoparticles by salicylaldiminepropyl results in efficient adsorbents for removal of Th⁴⁺, UO ₂ ²⁺ and Eu³⁺ ions from aqueous solutions. The effect of parameters influencing the adsorption efficiency such as aqueous phase pH, contact time, initial metal ions concentration, adsorbent dosage and temperature dependency of the process was verified and discussed. Under optimal conditions (pH 5.5, adsorbent dosage 0.05 g, contact time 30 min. and 25 °C), thorium and uranyl ions (initial concentration 20 mg/l) were quantitatively removed from 20 ml of sample solution. Under such conditions 85% of europium ions was removed. Comparison of the adsorption efficiency of the studied modified nano-particles with those unmodified ones shows a shift for uptake of the metal ions vs. pH curves towards lower pH values by applying the modified adsorbents. In addition, a significant improvement of europium ions adsorption was observed by using the modified nanoparticles. Kinetics of the process was studied by considering a pseudo second-order model. This model predicts chemisorption for the adsorption mechanism. Freundlich, Langmuir and Temkin models were suitable for describing the equilibrium data of Th⁴⁺, UO₂ ²⁺ and Eu³⁺ adsorption process, respectively. Thermodynamic investigation reveals the adsorption process of the studied ions is entropy driven.     

Development of brannerite glass‐ceramics for the immobilization of actinide‐rich radioactive wastes

Brannerite‐based glass‐ceramics have been developed as potential waste forms for the immobilization of actinide‐rich radioactive wastes. For the first time, the formation of brannerite phases in glass has been demonstrated using uranium (U) and plutonium (Pu) with additions of gadolinium and hafnium as neutron absorbers. Both XRD and SEM‐EDS confirm that brannerite is the dominating phase with compositions close to Y_0.5U_0.5Ti₂O₆, Gd_0.2Pu_0.3U_0.5Ti₂O₆, and Gd_0.1Hf_0.1Pu_0.2U_0.6Ti₂O₆ internally crystallized in the glass. TEM SAED and Raman spectroscopy reveal the typical structure and vibration modes for brannerite. In addition, the presence of U⁵⁺ species as designed in the formulations has been confirmed by diffuse reflectance spectroscopy. More importantly, the U and Pu were partitioned exclusively in the ceramic phases with no detectable actinide in the glass.     

Back-extraction Of Pu4 + From 20% Tributyl Phosphate – 20% Mixed Trialkyl Phosphine Oxides/kerosene In The Presence Of

Back-extraction of Pu^4 +  from a mixture of 20% tributyl phosphate (TBP) and 20% mixed trialkyl phosphine oxides (TRPO) in kerosene in the presence of UO²⁺ ₂was studied. The back-extractants investigated may be divided into three groups: carboxylic acids and salts, amino polycarboxylates, and phosphonic acid. The distribution coefficients of both Pu^4 +  and UO²⁺ ₂using a number of different back-extractants were measured and compared. The results obtained suggest that the only practical back-extractants are carboxylic acids. Among the carboxylic acids tested, oxalic acid is suitable when the UO²⁺ ₂ concentration in the organic phase is less than 2 g/L. For UO²⁺ ₂concentrations between 2 and 10 g/L, oxalic acid-nitric acid mixtures may be used. For UO²⁺ ₂concentrations greater than 10 g/L, the only practical back-extractant is glycolic acid. The results obtained here may be used to further develop a new process for separation of Pu^4 +  and UO²⁺ ₂ from TBP-TRPO/kerosene mixture.     

Extraction of Uranium(VI) and Plutonium(IV) with Tetra-Alkylcarbamides

ABSTRACT

The use of tetra-alkylcarbamides as novel extractants for the separation of uranium(VI) and plutonium(IV) by solvent extraction from spent nuclear fuels is investigated in this study. Batch extraction experiments show that tetra-alkylcarbamides strongly extract U(VI) with high distribution ratios. Plutonium(IV) can be co-extracted with U(VI) at high nitric acid concentration, while high U(VI)/Pu(IV) selectivities can be reached at lower acidity. Loading capacity experiments with high uranium concentrations show that alkyl chains longer than butyl are necessary to avoid third phase formation. Nevertheless, the viscosity of uranium-loaded solvents gets too high with alkyl chains longer than pentyl. Overall, this study shows that with TPU extractant (with four pentyl chains), an efficient co-extraction of uranium and plutonium can be reached (D_U,Pu > 1) for a concentration of nitric acid higher than 4 mol?L^?1, while the partition between uranium(VI) and plutonium(IV) could be operated even at 2 mol?L^?1 nitric acid without redox chemistry.     

ION EXCHANGE BEHAVIOUR OF CHELATING RESIN DOWEX A-1WITH ACTINIDES AND LANTHANIDES

Adsorption studies of several actinides and lanthanides have been carried out by chelating ion exchange resin Dowex A-1. The metal ions studied were Pu⁴⁺, Zr⁴⁺, UO₂ ⁺⁺, Am³⁺, Cm³⁺, Bk³⁺, Cf³⁺, Eu³⁺, and Tm³⁺. The separation factors between consecutive trivalent actinides and between Am(III) and Eu(III) have been evaluated. Mechanism of adsorption of actinides and lanthanides from different aqueous media has been discussed. An ion exchange procedure for the separation of Pu⁴⁺ and UO₂ ⁺⁺ has been developed using this resin.     

SOLVENT EXTRACTION OF URANIUM(VI) AND THORIUM(IV) WITH A TETRA-CARBOXYLATED CALIX[4]ARENE AND EFFECT OF ALKALI IONS (Na+, K+)

ABSTRACT

The solvent extraction of uranium(VI) and thorium(IV) with a tetra-carboxylated calix[4]arene (LH₄) in chloroform has been studied in the presence or absence of alkali ions (M⁺=Na⁺, K⁺). When studied alone, UO₂ ²⁺ and Th⁴⁺ were extracted into chloroform as 2:2 and 1:1 metal:ligand complexes, respectively. The efficiency of extraction increases in the presence of alkali ions, due to the formation of heteronuclear complexes. For uranium(VI), the extracted species are found to be both 2:2:2 and 1:1:1 (uoi₂ ²⁺:M⁺:LH₄,) mixed complexes. For Th(IV) in the presence of Na⁺, the formation of a mixed complex in 1:1:1 (Th(IV):Na⁺:LH₄,) proportions has been evidenced. However, the exact nature of this species could not be determined. In practical grounds, LH. may be useful as a selective extracting agent for Th(IV) with respect to U(VI) since separation factor Th(IV)/U(VI) close to 1000 have been measured in competitive extraction, in the presence or absence of alkali ions.     

Studies on the Development of a Flow-Sheet for AHWR Spent Fuel Reprocessing Using TBP

The present paper describes the results of solvent extraction studies carried out in batch mode to collect data on distribution of uranium, plutonium, and thorium using 5% TBP in n-dodecane. Extraction studies are carried out from feed solutions having bulk thorium containing aluminum and fluoride ions in ～3.00–4.00 M nitric acid at concentration levels anticipated in feed solutions during Advanced Heavy Water Reactor (AHWR) spent fuel reprocessing. Studies are carried out under varied experimental conditions. Parameters such as organic to aqueous phase ratio during extraction, concentration of nitric acid for scrubbing co-extracted thorium from loaded organic phase etc., are studied in detail. Hydroxylamine nitrate is selected for reductive stripping of plutonium in preliminary studies. Reagent mixture containing 0.30 M HAN + 0.60 M HNO₃ and 0.20 M N₂H₄ is found to be optimum for plutonium partitioning. This paper also describes the extraction and stripping of uranium and plutonium in co-current mode. The extraction behavior of relevant fission products is studied from a simulated feed solution. A preliminary study on a few commercially available reducing agents is also included. These data are useful in developing a flow-scheme for the recovery of uranium and plutonium from spent fuel originating from AHWR.     

Glycolamide-functionalized ionic liquid: Synthesis and actinide ion extraction studies

A glycolamide-functionalized ionic liquid (G-FIL) was synthesized for the first time and was evaluated for the extraction of actinide ions such as Am³⁺, Pu⁴⁺ and UO₂²⁺ and fission product element ions such as Eu³⁺, Sr²⁺ and Cs⁺. The extraction of the trivalent metal ions was found to be exceptionally high at low acid concentrations, which rapidly decreased with increasing acidity. In view of the high viscosity of the G-FIL, the studies were carried out using its diluted solution in a commercial ionic liquid, viz. 1-butyl-3-methylimidazolium bis(trifluoromethylsulfonyl)imide ([C₄mim][Tf₂N]).     

Synthesis,structure, and characterization of the thorium zirconolite CaZr1-xThxTi2O7 system

A series of zirconolite ceramics with composition CaZr_1-xTh_xTi₂O₇ (Δx = 0.10) were reactively sintered at 1350°C for 20 h, in air (0 ≤ x ≤ 0.60) and 5% H₂/N₂ (0 ≤ x ≤ 0.40). A sample with composition corresponding to x = 0.20 was also produced by hot isostatic pressing (HIP) at 1300°C and 100 MPa for 4 hours. Th⁴⁺ immobilization was most readily achieved under oxidizing conditions, with Th⁴⁺ preferentially incorporated within a pyrochlore-structured phase in the range 0.10 ≤ x ≤ 0.50, yet formation of the zirconolite-4M polytype was not observed. We report the novel synthesis of single-phase pyrochlore with nominal composition CaZr_0.40Th_0.60Ti₂O₇ when targeting x = 0.60. Th⁴⁺ incorporation under reducing conditions produced a secondary Th-bearing perovskite, comprising 24.2 ± 0.6 wt% of the phase assemblage when targeting x = 0.40, alongside 8.8 ± 0.3 wt% undigested ThO₂. Under reducing conditions, powder XRD data were consistent with zirconolite adopting the 3T polytype structure. The sample produced by HIP presented a nonequilibrium phase assemblage, yielding a major phase of zirconolite-2M alongside accessory Th⁴⁺-bearing phases ThTi₂O₆, ThO₂, and perovskite. These data highlight the efficacy of Th⁴⁺ as a Pu⁴⁺ surrogate, with implications for the formation of Zr-stabilized Th-pyrochlore phases as matrices for waste with elevated Th⁴⁺ content.