On the Basic Extraction Properties of a Phenyl Trifluoromethyl Sulfone-Based GANEX System Containing CyMe4-BTBP and TBP

A Grouped ActiNide EXtraction (GANEX) process for the extraction of actinides from used nuclear fuel for transmutation purposes has been investigated. The studied solvent consists of phenyl trifluoromethyl sulfone (FS-13), CyMe₄-BTBP, and TBP, a combination that has previously shown promising results. The time to reach extraction equilibrium for the system has been found to be less than 20 min. A 2:1 complex has been found between CyMe₄-BTBP and americium(III) or curium(III), whereas plutonium(IV) and CyMe₄-BTBP create a 1:1 complex. The extraction of fission product is low in the system.     

Direct Selective Extraction of Trivalent Americium from PUREX Raffinate Using a Combination of CyMe4BTPhen and TEDGA—A Feasibility Study

The direct and selective extraction of Am(III) from simulated PUREX raffinate is demonstrated using a novel combination of the lipophilic extractant CyMe₄BTPhen (2,9-bis(5,5,8,8-tetramethyl-5,6,7,8-tetrahydrobenzo[e]-[1,2,4]triazin-3-yl)-1,10-phenanthroline) and the hydrophilic complexant TEDGA (N,N,N’,N’-tetraethyl-diglycolamide) to enhance selectivity toward Am(III) extraction. Separation factors (SF) of up to SF_Am/Cm = 4.9 were observed in tracer experiments using this combination of CyMe₄BTPhen and TEDGA. Distribution ratios of stable isotopes of fission and activation products contained in a simulated PUREX raffinate solution are reported for the first time with CyMe₄BTPhen, and some co-extracted metal ions are identified. The metal ions partly co-extracted from the simulated PUREX raffinate solution were Cu, Pd, Cd, Ag, Ni, and to a lesser extent Sn and Mo. The co-extraction of Pd and Ag was successfully suppressed using Bimet ((2S,2’S)-4,4’-(ethane-1,2-diylbis(sulfanediyl))bis(2-aminobutanoic acid)). The extraction was also studied as a function of the TEDGA concentration. The distribution ratios of Am and Cm can be adjusted by variation of the TEDGA concentration to yield D_Am values >1 and D_Cm values <1. Separation factors for Am(III) over Cm(III) of up to SF_Am/Cm = 2.4 were observed in these experiments. For Ln(III) + Y(III), distribution ratios below 1 were observed, thus enabling a direct extraction of Am(III) from simulated PUREX raffinate with a sufficient selectivity against trivalent lanthanides and Cm(III).     

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.     

A Comparison of Americium Extractions as a Function of Time using Two Bis-Triazine-Bipyridine Ligands in Long-Chained Alcohol Diluents

The rate of americium mass transfer between the aqueous and organic phase in a solvent extraction system has been investigated. The ligand used is CyMe₄-BTBP and the diluents are long-chained alcohols. The results are compared with earlier reported data using C5-BTBP as ligand. In the C5-BTBP system the rate of the extraction could be correlated with the interfacial tension of the system, while not in the CyMe₄-BTBP system. In the CyMe₄-BTBP system a high (> 12) or low (< 9) dielectric constant of the diluent affects the equilibrium distribution ratio. Dielectric constants in between these two did not affect the extraction.     

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.     

Synthesis and Screening of t-Bu-CyMe4-BTBP,and Comparison with CyMe4-BTBP

The effect of adding a t-butyl group to the core molecule of CyMe₄-BTBP, with the aim of improving solubility in organic diluents, has been studied with regard to the extraction of Am(III) and Eu(III) from HNO₃. Synthesis of t-Bu-CyMe₄-BTBP is described in detail. Metal nitrates are extracted from nitric acid in the form of 1:2 complexes, M(NO₃)₃(BTBP)₂. Whether in 1-octanol, kerosene, or cyclohexanone diluents, t-Bu-CyMe₄-BTBP extracts with larger distribution ratios but with slower kinetics than CyMe₄-BTBP. The general trends previously observed for CyMe₄-BTBP regarding the diluent and modifier influence were also found for t-Bu-CyMe₄-BTBP.     

A TBP/BTBP-Based GANEX Separation Process – Part 3: Fission Product Handling

A Group ActiNide EXtraction (GANEX) separation system for transmutation has been developed, combining CyMe₄-BTBP with TBP and cyclohexanone. This new GANEX solvent has proven efficient in actinide extraction but also been found to extract some undesired fission products and corrosion products. Three major fission products were primarily selected for the study: Mo, Zr, and Pd. There are three main strategies for handling the extraction problem, all of which have been investigated and discussed; these are Pre-extraction, Suppression, and Scrubbing. The only strategy that was found to control the behavior of all three main fission products was suppression by the combination of two water-soluble complexing agents bimet and mannitol.     

Demonstration of a TODGA based Extraction Process for the Partitioning of Minor Actinides from a PUREX Raffinate

Abstract: Efficient recovery of minor actinides (MA) from genuine PUREX raffinate has been successfully demonstrated by the TODGA + TBP extractant mixture dissolved in an industrial aliphatic solvent TPH. The process was carried out in centrifugal contactors using an optimized flow‐sheet involving a total of 32 stages, divided into 4 stages for extraction, 12 stages for scrubbing and 16 stages for back‐extraction. Very high feed decontamination factors were obtained (Am, Cm ～ 40 000) and the recovery of these elements was higher than 99.99%. Of the non‐lanthanide fission products only Y and a small part of Ru were co‐separated into the product fraction together with the lanthanides and the MA.     

Recovery of Plutonium and Americium from Chloride Salt Wastes by Solvent Extraction

Abstract

Plutonium and americium can be recovered from aqueous waste solutions containing a mixture of HCl and chloride salt wastes by the coupling of two solvent extraction systems: tributyl phosphate (TBP) in tetra-chloroethylene (TCE) and octyl(phenyl)-N, N-diisobutyl carbamoylmethylphosphine oxide (CMPO) in TCE. In the flowsheet developed, the salt wastes are dissolved in HC1, the Pu(III) is oxidized to the IV state with NaC102 and recovered in the TBP-TCE cycle, and the Am is then removed from the resultant raffinate by the CMPO-TCE cycle. The consequences of the feed solution composition and extraction behavior of these species on the process flowsheet design, the Pu-product purity, and the decontamination of the aqueous raffinate from transuranic elements are discussed.     

Development of a Solvent Extraction Model for Process Tests in Short Residence Time Centrifugal Contactors

A computer program has been developed for optimization and modelling of counter–current solvent extraction processes. The distribution between the phases is calculated by either D-ratio functions or by a novel kinetic model for the transfer between the phases. The kinetic model is important to use when slow extraction kinetics yields D-ratios far from equilibrium. Transfer rate data was investigated in a single stage centrifugal contactor, modified for internal recirculation of the phases. Using this methodology a demonstration process for the recovery of minor actinides in a counter–current centrifugal contactor system using CyMe₄-BTBP was modelled with excellent agreement towards the experimental values.     

Demonstration of a TODGA‐Based Continuous Counter‐Current Extraction Process for the Partitioning of Actinides from a Simulated PUREX Raffinate,Part II: Centrifugal Contactor Runs

Abstract

The efficiency of the partitioning of trivalent actinides from a PUREX raffinate is demonstrated with a TODGA+TBP extractant mixture dissolved in an industrial aliphatic solvent TPH. Based on the results of cold and hot batch extraction studies and with the aid of computer code calculations, a continuous counter‐current process is developed and two flowsheets are tested using miniature centrifugal contactors. The feed solution used is a synthetic PUREX raffinate, spiked with ²⁴¹Am, ²⁴⁴Cm, ²⁵²Cf, ¹⁵²Eu, and ¹³⁴Cs. More than 99.9% of the trivalent actinides and lanthanides are extracted and back‐extracted and very high decontamination factors are obtained for most fission products. The co‐extraction of zirconium, molybdenum, and palladium is prevented using oxalic acid and HEDTA. However, 10% of ruthenium is extracted and only 3% is back‐extracted using diluted nitric acid. The experimental steady‐state concentration profiles of important solutes are determined and compared with model calculations and good agreement is generally obtained.     

Extracting Acetic Acid from Acidic Solutions

Abstract

In the UREX + process, acetic acid must be removed from the raffinate stream to avoid interference with the recovery and recycle of nitric acid solutions. Solvent extraction was selected to be the most promising approach to accomplish this cleanup. Acetic acid partitioning into pure diluents used in the UREX + process were found to be too low for an effective separation. Of the solvents tested, the most promising solvents for the extraction of acetic acid were found to be TBP in dodecane and TBP in FS-13.     

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.     

Solvent Extraction and Spectrophotometric Studies on Synergistic Extraction of Plutonium(IV) by Mixtures of Thenoyltrifluoroacetone (HTTA) and Tri-n-butylphosphate (TBP) in Benzene

Abstract

The synergistic extraction of Pu(IV) from perchloric acid solutions into mixtures of thenoyltrifluoroacetone (HTTA) and tri-n-butylphosphate (TBP) in benzene was investigated by solvent extraction methods. The adduct responsible for synergism was found to be Pu(TTA)₄·TBP. The adduct formation between Pu(TTA)₄ and TBP in the benzene phase was also investigated by spectrophotometry. The equilibrium constants for the equilibria involved were obtained both by solvent extraction and by spectrophotometric methods.     

Laboratory-Scale Counter-Current Centrifugal Contactor Demonstration of an Innovative-SANEX Process Using a Water Soluble BTP

In this paper the development and laboratory-scale demonstration of a novel “innovative-SANEX” (Selective Actinide Extraction) process using annular centrifugal contactors is presented. In this strategy, a solvent comprising the N,N,N’,N’-tetraoctyldiglycolamide (TODGA) extractant with addition of 5 vol.-% 1-octanol showed very good extraction efficiency of Am(III) and Cm(III) together with the trivalent lanthanides (Ln(III)) from simulated Plutonium Uranium Refining by Extraction (PUREX) raffinate solution without 3^rd phase formation. Cyclohexanediaminetetraacetic acid (CDTA) was used as masking agent to prevent the co-extraction of Zr and Pd. An(III) and Ln(III) were co-extracted from simulated PUREX raffinate, and the loaded solvent was subjected to several stripping steps. The An(III) were selectively stripped using the hydrophilic complexing agent SO₃-Ph-BTP (2,6-bis(5,6-di(sulfophenyl)-1,2,4-triazin-3-yl)pyridine). For the subsequent stripping of the Ln(III), a citric acid based solution was used. A 32-stage process flow-sheet was designed using computer-code calculations and tested in annular miniature centrifugal contactors in counter-current mode. The innovative SANEX process showed excellent performance for the recovery of An(III) from simulated High Active Raffinate (HAR) solution and separation from the fission and activation products. ≥ 99.8% An(III) were recovered with only low impurities (0.4% Ru, 0.3% Sr, 0.1% Ln(III)). The separation from the Ln(III) was excellent and the Ln(III) were efficiently stripped by the citrate-based stripping solution. The only major contaminant in the spent solvent was Ru, with 14.7% of the initial amount being found in the spent solvent. Solvent cleaning and recycling therefore has to be further investigated. This successful spiked test demonstrated the possibility of separating An(III) directly from HAR solution in a single cycle which is a great improvement over the former multi-cycle strategy. The results of this test are presented and discussed.     

Water Content of Organic Solvents and their Relationship to Extraction of Nitric and Acetic Acid from UREX+ Streams

This study investigates the equilibrium absorption of water in various solvents and solvent-mixtures being considered for the counter-current solvent extraction of acetic acid from improved Uranium Extraction (UREX+) process solutions. It then seeks to determine if there is any correlation between the equilibrium water content of these solvents and their equilibrium extraction of 0.25 M nitric and 0.025 M acetic acid. The UREX+ process is a proliferation resistant version of the Plutonium Uranium Extraction (PUREX) process. The solvents studied were n-Dodecane (nDD), 1,2 Dichloroethane (DCE), and Phenyltrifluoromethyl Sulfone (FS-13), and mixtures of these solvents with Tributyl Phosphate (TBP). After studying both pure water and acidified aqueous systems, it seems the water absorption mechanism is independent of the diluent used and remains constant with the addition of the 0.25 M nitric and 0.025 M acetic acid.     

THE TRUEX PROCESS - A PROCESS FOR THE EXTRACTION OF THE TKANSURANIC ELEMENTS EROM NITRIC AC In WASTES UTILIZING MODIFIED PUREX SOLVENT*

ABSTRACT

A generic transurantc (TRU) element extraction/recovery process was developed based on the use of octyl(phenyl)-N,N-diiso-butylcarbamoylmetliylphosphine oxide, 0φD(iB)CMPO, dissolved in PUREX process solvent (tribntyl phosphate, TBP, in normal paraffluic hydrocarbon, NPH). The process (called TRUEX) is capable of reducing the TRU concentration by many orders of magnitude In waste solutions containing a wide range of nitric acid, salt, and fission product concentrations. A major feature of the process is that it is readily adaptable for waste processing in existing fuel reprocessing facilities.     

DISTRIBUTION OF N- AND ISO- BUTYRALDEHYDES BETWEEN TRI-N-BUTYL PHOSPHATE/ N-DODECANE AND NITRIC ACID

ABSTRACT

The distribution of n- and iso-butyraldehydes between tri-n-butyl phosphate(TBP) n-dodecane(nDD) and HNO₃ were measured. The distribution ratio of n-butyraldehyde in the TBP/nDD and HNO₃ system was nearly the same as that of iso-butyraldehyde. The distribution ratios of n- and iso-butyraldehydes increased with TBP concentration in the organic phase. The equilibrium constant of the extraction reaction was about 2. In a uranium, neptunium and plutonium separation process, most of the n- and iso-butyra1dehydes fed into theNp separation stepor into thePu/U partition will be left with the TBP solvent. The two compounds will be partly back-extracted to the aqueous phase in the U purification and in the solvent washing steps of the PUREX process.     

Study of the Mass Transfer Behavior in a Centrifugal Contactor and Verification of the Solvent Extraction Model for the SX Process Program

The SX Process program has been developed for modelling of extraction processes in centrifugal contactors where the transfer kinetics is of big importance due to the short hold up time. Apparent distribution ratios are calculated using a stage efficiency which is flow-rate independent. In this work the dependency of the stage efficiency on parameters affecting the extraction transfer rate, such as metal loading, O/A ratio and acidity, has been investigated in single stage centrifugal contactor experiments for extraction of americium(III) into a 0.015 M CyMe₄-BTBP/0.25 M DMDOHEMA/octanol system. A model is proposed on how to calculate the stage efficiency and to accurately predict the apparent distribution ratios under the different conditions used.     

EXTRACTION STUDY FOR TRUEX PROCESS USING SHORT-LIVED RADIONUCLIDES PRODUCED BY NEUTRON IRRADIATION OF URANIUM

The extraction behavior of short-lived fission products and neptunium was studied by using octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide under the conditions of the transuranic elements extraction (TRUEX) process. The short-lived fission products and neptunium were produced by neutron irradiation of UO₂ of natural uranium, and the extraction behavior of ⁹³Y, ⁹⁹Mo, ⁹⁷Zr, ¹²²Sb, ¹³²Te, ¹³³I, ¹⁴³Ce, and ²³⁹Np was simultaneously studied, where ¹²²Sb was produced by neutron irradiation of antimony metal. The extraction of fission products and Np under the conditions of the PUREX process was also studied for comparison. The extraction of nuclides in the presence of large amounts of uranium(VI), and the presence of oxalic acid was examined. The conditions and performance of the TRUEX extraction were discussed by considering the obtained results.