A. S. Kanekar

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, DPu 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 DPu 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.

Comparison of Hydrometallurgical Parameters of N,N‐Dialkylamides and of Tri‐n‐Butylphosphate

Straight‐chain N,N‐dihexyloctanamide (DHOA) and branched‐chain N,N‐di(2‐ethylhexyl)isobutyramide (D2EHIBA) have been identified as promising alternatives to tri‐n‐butylphosphate (TBP) for the reprocessing of spent uranium based fuels, and selective extraction of 233U from irradiated thorium fuels, respectively. The present work deals with the effects of different hydrodynamic parameters such as viscosity, density, and interfacial tension (IFT) on the phase‐separation time (PST) under uranium and thorium loading conditions. The IFT values have been determined under varying experimental conditions such as the aqueous nitric acid concentration, n‐dodecane purity, ligand concentration, and thorium/uranium loading conditions. These studies have suggested that the quality of n‐dodecane affects the IFT values of different solutions. The IFT values of D2EHIBA changed marginally (23.3 ± 0.9 mNm−1) against THOREX feed solution for the wide range of D2EHIBA concentration (0.1–1.0 M). However, IFT, viscosity, and PST values increased with uranium loading of 1.1 M DHOA. These studies suggested that a lower phase‐disengagement rate with increased uranium loading was mainly due to the increased viscosity of the loaded 1.1 M DHOA solution.

Evaluation of N,N-dihexyloctanamide as an alternative extractant for spent fuel reprocessing: batch and mixer settler studies

ABSTRACT Solvent extraction studies carried out at BARC, India on the evaluation of N,N-dialkyl amides as alternative extractants to tri-n-butyl phosphate (TBP) for reprocessing of spent fuel have suggested that straight chain N,N-dihexyloctanamide (DHOA) is promising alternative to TBP for the reprocessing of irradiated uranium based fuels. This paper presents the batch as well as mixer settler studies for uranium and plutonium extraction/stripping to evaluate DHOA vis a-vis TBP for the reprocessing of Pu rich fuels. These studies showed that uranium extraction using DHOA as extractant was comparable to that of TBP; however, it displayed better stripping behavior than TBP. Plutonium extraction behavior was better in the case of DHOA as compared to that of TBP. However, Pu stripping data indicated towards the need of reducing agent in the stripping cycle for both the extractants.

A rapid online estimation method for radiostrontium in soil samples using crown ether and supercritical fluid extraction.

Crown ethers dissolved in suitable medium are well known to promote the extraction of alkali (M(+)) and alkaline-earth (M(2+)) cations from aqueous to organic phases. Di-tert-butyl-cyclohexano18crown6 (DTBDCH18C6) has been identified as an effective and selective extractant for Sr(II) from nitric acid medium. An attempt was made to evaluate the feasibility of (85,89)Sr recovery from synthetic soil samples (0.5 g; particle size: <100 μm) by SFE route (pressure: 200 kg/cm(2); T: 40 °C) employing DTBDCH18C6 dissolved in methanol/nitric acid medium as phase modifier. The effect of various experimental parameters such as (i) dynamic/static mode of extraction, (ii) time of equilibration (15-150 min during static mode of extraction using 3 mL of modifier), (iii) nitric acid concentration (1-6M), (iv) picrate as counter-anion, and (v) crown ether concentration in the modifier phase (2×10(-4)-2×10(-3) M) on Sr(II) extraction was studied. Based on these studies, 2×10(-4) M DTBDCH18C6 dissolved in methanol/4M HNO(3) was chosen as modifier and 30 min as equilibration time for batch mode employing 3 mL modifier solution in the static mode. Three successive batches employing 3 mL modifier solution (after each extraction stage) showed near quantitative recovery (>95%) of (85,89)Sr from soil samples. Dynamic mode extraction using 2×10(-4) M DTBDCH18C6 dissolved in methanol/4M HNO(3) as modifier suggested that near quantitative recovery (>95%) of (85,89)Sr could be achieved within 1h. By contrast, ~10% (137)Cs extraction was observed from soil samples under identical experimental conditions. These studies demonstrate the potential of the SFE technique for the analysis of (90)Sr in different environmental samples.

Supercritical fluid extraction of uranium from sintered oxides (UO2, (U,Th)O2), soil and ore samples using tri-n-butylphosphate and N,N-di-(2-ethylhexyl) isobutyramide as extractants

ABSTRACT Direct extraction of uranium from different samples viz. sintered UO2, (U,Th)O2, soil, and ores was carried out by modifier free supercritical fluid containing tri-n-butylphosphate (TBP) and/or N,N-di-(2-ethylhexyl)isobutyramide (D2EHIBA) as extractants. These extractants were pre-equilibrated with nitric acid prior to their use in supercritical fluid extraction experiments. Uranium extraction studies from sintered UO2 showed that pre-equilibration with more concentrated nitric acid helped in its better dissolution and extraction. The extraction of uranium from (U,Th)O2 samples was signifi cantly lower for both TBP–HNO3 (∼17%) and D2EHIBA–HNO3 (∼12%) adducts in 2 h, under the conditions of present study. Modifier free supercritical fluid extraction appears attractive with respect to minimization of secondary wastes. This method resulted 80–100% extraction of uranium from different soil/ore samples. The results were confirmed by performing neutron activation analysis of original (before extraction...

Evaluation of N,N-dialkylamides as promising process extractants

Studies carried out at BARC, India on the development of new extractants for reprocessing of spent fuel suggested that while straight chain N,N-dihexyloctanamide (DHOA) is promising alternative to TBP for the reprocessing of irradiated uranium based fuels, branched chain N,N-di(2-ethylhexyl)isobutyramide (D2EHIBA) is suitable for the selective recovery of 233U from irradiated Th. In advanced fuel cycle scenarios, the coprocessing of U/Pu stream appears attractive particularly with respect to development of proliferation resistant technologies. DHOA extracted Pu(IV) more efficiently than TBP, both at trace-level concentration as well as under uranium/plutonium loading conditions. Uranium extraction behavior of DHOA was however, similar to that of TBP during the extraction cycle. Stripping behavior of U and Pu (without any reductant) was better for DHOA than that of TBP. It was observed during batch studies that whereas 99% Pu is stripped in four stages in case of DHOA, only 89% Pu is stripped in case of TBP under identical experimental conditions. DHOA offered better fission product decontamination than that of TBP. GANEX (Group ActiNide EXtraction) and ARTIST (Amide-based Radio-resources Treatment with Interim Storage of Transuranics) processes proposed for actinide partitioning use branched chain amides for the selective extraction of uranium from spent fuel feed solutions. The branched-alkyl monoamide (BAMA) proposed to be used in ARTIST process is N,N-di-(2-ethylhexyl)butyramide (D2EHBA). In this context, the extraction behavior of U(VI) and Pu(IV) were compared using D2EHIBA, TBP, and D2EHBA under similar concentration of nitric acid (0.5 — 6M) and of uranium (0-50g/L). These studies suggested that D2EHIBA is a promising extractant for selective extraction of uranium over plutonium in process streams. Similarly, D2EHIBA offered distinctly better decontamination of 233U over Th and fission products under THOREX feed conditions. The possibility of simultaneous stripping and precipitation of thorium (as oxalate) from loaded organic phase was explored using 0.05M oxalic acid. Ammonium diuranate (ADU) precipitation was performed on the oxalate supernatant for the recovery of uranium. Quantitative recovery (>99.9%) of Th as well as of U was achieved. Radiolytic studies suggested that irradiated DHOA and D2EHIBA behaved better with respect to fission product decontamination as compared to that of TBP.

Supercritical Fluid Dissolution and Extraction of Trivalent Metal Cations from Different Matrices

Studies on the recovery of trivalent metal ions such as Nd3+Eu3+ (taken as homologs of Am(III)) from solid oxide (Nd2O3), Thorium concentrate (obtained from Monazite ore processing), tissue paper/surgical gloves (rubber), and plant samples have been carried out by supercritical fluid extraction (SFE) using supercritical CO2 and ethanol/nitric acid. N,N,N’,N’-tetraoctyl diglycolamide (TODGA) was used as the extractant in these studies. The results showed that the recovery of Nd increased with TODGA concentration from 50% (no TODGA) to 70% (10% TODGA) at 3 M HNO3 in ethanol. However, the extraction of Nd at 1 M HNO3 was invariant with 1-3% (v/v) TODGA concentration (73 ± 4%). Interestingly, REEs recovery from Th concentrate was ˜ 60% even without TODGA using ethanol/3 M HNO3 mixture. On the other hand, quantitative recovery of 152,154Eu from tissue paper and surgical gloves sample could be achieved using 3 M HNO3/ethanol mixture. This suggested that it would be possible to decontaminate the contaminated laboratory waste papers using SFE technique.