Actinide Partitioning with a Modified TODGA Solvent: Counter-Current Extraction Studies with Simulated High Level Waste

Actinide partitioning studies with improved N,N,N',N'-tetraoctyl diglycolamide (TODGA) solvent (0.05 M TODGA + 5% iso-decanol in n-dodecane) has been explored in order to achieve better decontamination from fission products. The distribution behavior of various metal ions, viz. Am, Pu, U, Eu, Sr, Pd, Cs, Tc, Fe, and Mo with improved TODGA solvent was investigated. Lower concentration of TODGA (0.05 M as compared to previously proposed 0.1 M or 0.2 M) exhibited required extraction properties for actinide partitioning from pressurized heavy water reactor high level waste (PHWR-HLW). Counter-current extraction studies with simulated PHWR-HLW spiked with different radio-tracers (viz. ²⁴¹Am, ¹⁵²Eu, ¹³⁷Cs, ^85,89Sr, ⁵⁹Fe, ¹⁰⁶Ru, ¹⁰⁹Pd, ⁹⁵Zr, and ⁹⁹Mo) suggested that > 99.9% of the trivalent actinides and lanthanides could be extracted in six stages and stripped in four stages. The decontamination factor for various fission products with respect to Am was: 3918 (Cs), 2990 (Sr), 1150 (Zr), 1407 (Ru), 1185 (Pd), and 3250 (Mo). The counter-current extraction studies with the irradiated solvent (500 kGy) reflected a significant amount of Mo extraction.     

Plutonium Loading of Prospective Grouped Actinide Extraction (GANEX) Solvent Systems based on Diglycolamide Extractants

The Grouped Actinide Extraction (GANEX) process is being developed for actinide recycling within future nuclear fuel cycles. Interactions between potential solvents and macro-concentrations of plutonium are one of the most important issues in defining the GANEX process. Surprisingly, plutonium loading of diglycolamide (DGA) based solvents such as tetra-octyl DGA (TODGA) causes precipitation rather than a conventional third phase, in direct contrast to results with U(VI), Th(IV) or lanthanide ions. Various DGA based solvent systems have been screened for their plutonium loading capacity and 0.2 M TODGA with 0.5 M DMDOHEMA in a kerosene diluent is selected as the optimum solvent formulation of those tested. Plutonium can be relatively easily stripped from this solvent using aqueous acetohydroxamic acid but this is very acid dependent in the low acidity region.     

The Effect of Alkyl Substituents on Actinide and Lanthanide Extraction by Diglycolamide Compounds

The effect of the alkyl substituents on amidic N atoms in diglycolamide (DGA) compounds on solvent extraction has been investigated. The solubility in water and n-dodecane, lanthanide loading capacity, and distribution ratios (D) of lanthanides and actinides for various DGA compounds are reported. DGA derivatives with short alkyl chains, for example, methyl and ethyl groups, are very water soluble, while DGA derivatives with long alkyl chains, for example, octyl (TODGA), decyl (TDDGA), dodecyl (TDdDGA), and 2-ethylhexyl (TEHDGA) group are moderately soluble in n-dodecane. DGA derivatives with phenyl substituents have very low solubility in both aqueous and organic solvents, which suggests that these compounds will not be suitable for solvent extraction applications in the HNO₃/n-dodecane systems. The lanthanide loading capacities of DGA extractants correlate with their alkyl chain lengths according to the following order: TDdDGA > TDDGA > TODGA > TEHDGA. The branched-alkyl-chain DGA derivative (TEHDGA) exhibits both lower D and loading capacity than TODGA. The results of masking-effect and solubility tests indicate that TEDGA is the best actinide masking agent among the water-soluble DGA derivatives tested. Actinide and lanthanide extractions using ten DGA compounds in six diluents (nitrobenzene, 1,2-dichloroethane, 1-octanol, chloroform, toluene, and n-dodecane) are also reported; it was observed that lipophilic DGA derivatives with shorter alkyl chains show higher D values.     

FF智能仪表在核燃料后处理试验萃取装置中的应用

介绍了一种基于FF智能仪表的核燃料后处理萃取装置的控制系统结构。针对萃取试验台架的测控要求,给出了FF智能仪表的应用和在Delta V系统中的组态方法。     

Decomposition of Solvent Extraction Media during Nuclear Reprocessing: Literature Review

Abstract

Reprocessing spent nuclear fuel is indispensable for the economical use of uranium in nuclear energy production, and has been used industrially for more than 40 years. These processes involve the use of an extractant/diluent (solvent) for separation of the reusable actinides from unwanted fission products. The most widely used processes employ tributyl phosphate (TBP) diluted with normal-paraffin hydrocarbon. However, on repeated use, the solvent becomes degraded due to thermal, radiolytic, and chemical attacks, resulting in chemical as well as physical damage. In view of the considerable expansion in the knowledge and understanding regarding the chemical and radiolytical decomposition of both TBP and the hydrocarbon diluent, an up-to-date review seemed appropriate. This review is concerned mainly with the mechanisms of the degradation of the solvent system. Schemes of TBP and diluent radiation-chemical transformations occurring on decomposition of the solvent system are presented.     

Non-Dispersive Solvent Extraction and Stripping of Neodymium (III) using a Hollow Fiber Contactor with TODGA as the Extractant

Membrane based non-dispersive solvent extraction (NDSX) of Nd(III) from aqueous nitric acid medium was carried out using a commercial liquid-liquid extraction system consisting of an extra-flow hollow fiber module with about 10,000 microporous hydrophobic polypropylene capillaries with an effective surface area of 1.4 m². The NDSX operation was carried out by pumping about 1 g/L Nd(III) solutions at a fixed nitric acid concentration of 3.57 M through the tube side and organic solvent (tetra-n-octyl-3-oxapentane-diamide (TODGA) in normal paraffin hydrocarbon (NPH)) through the shell side of the hollow fiber capillaries. NDSX studies were performed under different hydrodynamic conditions and the overall mass transfer was evaluated under counter-current flow conditions. Conditions for quantitative extraction and stripping were also identified. Reproducibility of the results from multiple runs was found to be excellent after five different identical runs.     

Solvent extraction behavior of trivalent metal ions in diglycolamides having same carbon to oxygen ratio

Various diglycolamides (DGAs), N,N,N’,N’-tetraoctyl diglycolamide (TODGA), N,N-di-2-ethylhexyl-N’,N’-di-octyl diglycolamide (DEHDODGA), N,N,N’,N’-tetra-2-ethylhexyl diglycolamide (TEHDGA), N,N-di-decyl-N’,N’-di-hexyl diglycolamide (D²DHDGA), N,N-di-butyl-N’,N’-di-dodecyl diglycolamide (DBD³DGA), N,N,N’,N’-tetra-hexyl diglycolamide (THDGA), and N,N,N’,N’-tetradecyl diglycolamide (TDDGA) have been synthesized and studied for the extraction of Am(III) and Eu(III) from nitric acid medium. An attempt was made to understand the extraction and third phase formation behavior of trivalent metal ions in these DGAs. Our results revealed that despite having the same carbon to oxygen ratio in these DGAs, the third phase formation behavior and extractions achieved in these DGAs are significantly different.     

Neptunium Extraction and Stability in the GANEX Solvent: 0.2 M TODGA/0.5 M DMDOHEMA/Kerosene

Batch distribution studies show that above ~1 M HNO₃ Np(IV) and Np(VI) are well extracted into a solvent of 0.2 M TODGA/0.5 M DMDOHEMA in kerosene that has been formulated for the extraction of transuranic actinides in the GANEX (grouped actinide extraction) process. Np(IV) and Np(VI) are quite stable in the solvent phase, although Np(VI) is slowly reduced to Np(IV) on standing. Stripping of Np(IV,VI) ions from the GANEX solvent has been shown to be quite efficient using acetohydroxamic acid providing aqueous HNO₃ concentrations are below ~0.3 M. In contrast, Np(V) shows much lower distribution ratios and is unstable in the GANEX solvent phase with quite rapid formation of Np(IV) observed. Closer analysis shows this to be due to Np(V) disproportionation, which is enhanced at higher organic phase acidities. Disproportionation of Np(V) is also shown to occur in separate TODGA and DMDOHEMA kerosene solutions. These observations thus enable conditions for neptunium extraction to be optimized during the design of a GANEX flowsheet.     

Development of a TODGA based Process for Partitioning of Actinides from a PUREX Raffinate Part I: Batch Extraction Optimization Studies and Stability Tests

Abstract

The extractant, N,N,N′,N′‐tetraoctyl diglycolamide (TODGA) has been evaluated for the separation of actinides(III) and lanthanides(III) from a high active raffinate (HAR). The effect of oxalic acid and HEDTA complexant on the extraction of actinides(III), lanthanides(III), and important fission products (e.g. Mo, Pd, Sr, Zr, Ru etc.) from synthetic HAR has been studied with 0.2 mol/L TODGA in TPH. With an extractant mixture of TODGA and tributyl phosphate (TBP) the amount of oxalic acid can be reduced to less than 0.3 mol/L for the effective complexation of zirconium, whereas the distribution ratios of actinides(III) and lanthanides(III) are still high for the separation from HAR. Furthermore the maximum loading of lanthanides (e.g. Nd) can be significantly increased by adding TBP to the extractant. However, the extraction of oxalic acid and nitric acid also increased by the addition of TBP, which can lead to problems during back extraction of the loaded extractant. Extraction studies after radiolysis and hydrolysis reveal that the TODGA+TBP mixture is a sufficient stable extraction system suited for further process development studies.     

Direct Selective Extraction of Actinides (III) from PUREX Raffinate using a Mixture of CyMe4BTBP and TODGA as 1-cycle SANEX Solvent Part III: Demonstration of a Laboratory-Scale Counter-Current Centrifugal Contactor Process

The direct selective separation of the trivalent actinides americium and curium from a simulated Plutonium Uranium Refining by EXtraction (PUREX) raffinate solution by a continuous counter-current solvent extraction process using miniature annular centrifugal contactors was demonstrated on a laboratory scale. In a 32-stage spiked test (12 stages for extraction, 16 stages for scrubbing, and 4 stages for Am/Cm stripping), an extractant mixture of CyMe₄BTBP and TODGA in a TPH/1-octanol mixture was used. The co-extraction of some fission and corrosion product elements, such as zirconium and molybdenum, was prevented by using oxalic acid. Co-extracted palladium was selectively stripped using an L-cysteine scrubbing solution and the trivalent actinides were selectively stripped using a glycolic acid-based stripping solution. It was demonstrated that a selective extraction and high recovery of > 99.4% of the trivalent minor actinides was achieved with low contamination by fission and corrosion products. The product contained 99.8% of the initial americium and 99.4% of the initial curium content. The spent solvent still contained high concentrations of Cu, Cd, and Ni. The experimental steady-state concentration profiles of important solutes were determined and compared with those from computer-code calculations.     

Direct Selective Extraction of Actinides (III) from PUREX Raffinate using a Mixture of CyMe4BTBP and TODGA as 1-cycle SANEX Solvent

Abstract

Within the framework of our research activities related to the partitioning of spent nuclear-fuel solutions, the direct selective extraction of trivalent actinides from a simulated PUREX raffinate was studied using a mixture of CyMe₄BTBP and TODGA (1-cycle SANEX). The solvent showed a high selectivity for trivalent actinides with a high lanthanide separation factor. However, the coextraction of some fission product elements (Cu, Ni, Zr, Mo, Pd, Ag, and Cd) from a simulated PUREX raffinate was observed, with distribution ratios up to 30 (Cu). The extraction of Zr and Mo could be suppressed using oxalic acid but the use of the well-known Pd complexant N-(2-Hydroxyethyl)-ethylendiamin-N,N′,N′-triacetic acid (HEDTA) was unsuccessful. During screening experiments with different amino acids and derivatives, the sulfur-bearing amino acid L-Cysteine showed good complexation of Pd and prevented its extraction into the organic phase without influencing the extraction of the trivalent actinides Am (III) and Cm (III). The optimization studies included the influence of the L-Cysteine and HNO₃ concentration and the kinetics of the extraction. The development of a process-like extraction series showed very promising results in view of further optimizing the process. A strategy for a single-cycle process is proposed within this article.     

关于均相溶剂萃取在分离分析应用中的讨论

提出了"均相溶剂萃取"的概念,比较了均相溶剂萃取与传统溶剂萃取的异同点,简要讨论均相溶剂萃取在分析化学中的部分应用.均相溶剂萃取法作为一种新的溶剂萃取分离法应当在分析化学教学中引起关注,以此拓展学生的知识面和培养学生的创新思维和创新意识.     

Development of Steam Reforming for the Solidification of the Cesium and Strontium Separations Product from Advanced Aqueous Reprocessing of Spent Nuclear Fuel

Abstract

Steam reforming is one option currently being investigated for stabilization of the cesium/strontium strip products from spent fuel reprocessing solvent extraction processes because it can potentially destroy the nitrates and organics present in these aqueous, nitrate‐bearing solutions, while converting the cesium and strontium into leach resistant aluminosilicate minerals, such as pollucite. To produce pollucite and other mineral analogs of the alkaline metals, the feeds must be mixed with aluminosilicate compounds and thermally sintered or calcined to activate solid‐state crystal formation. Scoping tests completed indicated that the cesium/strontium in these organic and acid solutions can be converted into aluminosilicate materials using steam reforming.     

Comparative Evaluation of CMPO and Diglycolamide Based Solvent Systems for Actinide Partitioning in Mixer-Settler Runs Using Tracer-Spiked PHWR Simulated High Level Waste

The TRUEX solvent (0.2 M CMPO + 1.2 M TBP) was employed for countercurrent extraction studies with radiotracers spiked pressurized heavy water reactor simulated high level waste (PHWR-SHLW) employing a 12-stage of mixer-settler. The results of the mixer-settler runs with CMPO were compared with those obtained under identical conditions employing TODGA (N,N,N’,N’-tetraoctyl diglycolamide) and T2EHDGA (N,N,N’,N’-tetra-2-ethylhexyl diglycolamide) as the extractants. Even though the TRUEX solvent revealed quantitative extraction of trivalent actinides and lanthanides in 5 stages at O/A = 1, stripping of the extracted metal ions from the loaded organic phase was poor with dilute HNO₃ solution (0.2 M HNO₃). Quantitative stripping could not be achieved in 12 stages even when a complexing solution (0.1 M citric acid + 0.1 M HNO₃) was employed as the strippant. In contrast, the stripping from loaded TODGA and T2EHDGA solvents was possible in < 6 stages with 0.2 M HNO₃. The experimental results suggested that the performance of TRUEX solvent was inferior to the diglycolamide based extractants such as TODGA and T2EHDGA.     

Diamylamylphosphonate Solvent Extraction of Am(VI) from Nuclear Fuel Raffinate Simulant Solution

The separation of hexavalent americium from the lanthanides in simulated PUREX raffinate solution using 1 M diamylamylphosphonate in dodecane extraction was investigated. Hexavalent americium was prepared using room-temperature sodium bismuthate oxidation. Under these conditions the majority of the lanthanides were not oxidized and remained inextractable. A separation factor of ～50 was provided for americium from europium over the nitric acid concentration range 6–7 M. Cerium was the exception with oxidation to Ce^IV resulting in its co-extraction with Am^VI. However, since americium is readily reduced to Am^III it was easily stripped with a dilute acidic solution of hydrogen peroxide. Although hydrogen peroxide also reduces cerium, it does so slowly, and a selective americium strip was achieved, with a separation factor of as high as 35. Alternatively, since americium spontaneously reduced in the loaded organic phase, samples allowed to stand for 2 hours could be selectively stripped of americium by contact with 1 M HNO₃ containing no additional reagents. Further, the separation was demonstrated using solutions containing macro-amounts of cerium and americium. The implications for use in fuel cycle separations are discussed.     

溶剂萃取回收废弃锂离子电池中钴金属的研究

丙酮溶解分离出正极材料,2 mol/L H2SO4+30%H2O2浸出,水解净化除杂后,P507三级萃取,H2SO4反萃取回收废弃锂离子电池中的钴元素,优化了各步骤的操作参数,最终得到适用于生产锂钴氧化物的硫酸钴,钴的回收率达到95%。     

选择性溶剂萃取生产低凝柴油的溶剂筛选和溶剂选择性的表征

酮类、酯类、氯代烃类溶剂分别对减一线油和常三线油进行了选择性萃取生产低凝柴油的实验。溶剂乙酸乙酯的选择性好,可降低减一线油和常三线油冷滤点分别为18°C和16°C,同时对柴油燃烧性能的影响较小,是选择性萃取生产低凝柴油的最佳溶剂。比较了各溶剂的脱蜡率、脱出的各碳数正构烷烃的质量分布和脱蜡油的冷滤点,引入了由高碳数正构烷烃脱出量、脱蜡率和冷滤点降低值三个因子组成的综合评价参数。该参数能表征本工艺要求的选择性,并直观地指导对两种油样均优的萃取溶剂的选择。     

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.     

Re-evaluation of Radiation-Energy Transfer to an Extraction Solvent in a Minor-Actinide-Separation Process Based on Consideration of Radiation Permeability

ABSTRACT

Absorbed-dose estimation is essential for evaluation of the radiation tolerance of minor-actinide-separation processes. We propose a dose-evaluation method based on radiation permeability, with comparisons of heterogeneous structures seen in the solvent-extraction process, such as emulsions forming in the mixture of the organic and aqueous phases. A demonstration of radiation-energy-transfer simulation is performed with a focus on the minor-actinide-recovery process from high-level liquid waste with the aid of the Monte Carlo radiation-transport code PHITS. The simulation results indicate that the dose absorbed by the extraction solvent from alpha radiation depends upon the emulsion structure, and that from beta and gamma radiation depends upon the mixer-settler-apparatus size. Non-negligible contributions of well-permeable gamma rays were indicated in terms of the plant operation of the minor-actinide-separation process.     

Synergistic Solvent Extraction of Lanthanides(III) with Mixtures of 4-benzoyl-3-methyl-1-phenylpyrazol-5-one and Some Novel Carbamoyl- and Phosphorylmethoxymethylphosphine Oxides

The solvent extraction of La, Ce, Pr, Nd, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, and Y from weak acidic hydrochloric acid solutions into an organic phase containing 4-benzoyl-3-methyl-1-phenylpyrazol-5-one (HP) and neutral tridentate organophosphorus ligands R₂P(O)CH₂OCH₂C(O)NBu₂ R = Bu (I), R = Ph (II) and R₂P(O)CH₂OCH₂P(O)R¹₂ R = R¹ = Bu (III); R = Bu, R¹ = Ph (IV); R = R¹ = Ph(V) has been studied. A considerable synergistic effect was observed in the presence of HP in the organic phase containing tetraoctyldiglycolamide (TODGA) and neutral organophosphorus ligands I - V. A successive replacement of C(O)NAlk₂ groups in the diglycolamide extractant molecule by P(O)Ph₂ groups leads to an increase in the extraction efficiency of Ln(III) ions when toluene was used as diluent. Phosphoryl-containing podand I possess a higher extraction efficiency towards Ln(III) ions than TODGA. The extraction equilibrium was investigated and the equilibrium constants were calculated. It was found that the lanthanide(III) ions are extracted as LnLP₃ and LnL₂P₃ complexes with mixtures of HP and I in toluene from weak acidic solutions.