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The extraction behavior of U(VI), Pu(IV), and Np(IV) from nitric acid medium has been studied using branched chain di(2-ethylhexyl)isobutyramide (D2EHiBA) dissolved in different room temperature ionic liquids (RTILs) [Cnmim][NTf2] (where n = 4, 6, or 8). Uranium extraction (DU) increased gradually with aqueous phase acidity for the three RTILs used in this study suggesting solvation mechanism. There was a reversal in the extraction behavior of Pu(IV) and Np(IV) from nitric acid medium using D2EHiBA dissolved in RTILs as solvents as compared to the behavior reported in the molecular diluent,n-dodecane, which shows negligible extraction of these metal ions. The extraction of Pu(IV) increased with aqueous phase acidity in different RTILs in the order: [C8mim][NTf2] > [C6mim][NTf2] > [C4mim][NTf2]. The distribution ratio values of these metal ions followed the order: DPu(IV)≥ DNp(IV)> DU(VI)using D2EHiBA as extractant suggesting that RTILs can modify the extraction behavior of extractants. [ABSTRACT FROM PUBLISHER]

Straight chainN,N-dihexyloctanamide (DHOA) has been identified as a promising alternative of tributyl phosphate (TBP) in the reprocessing of irradiated uranium based fuels. This article deals with the evaluation of DHOA vis-à-vis TBP as extractant in counter-current runs employing centrifugal contactors for uranium extraction from feed solutions of varying compositions such as uranium solutions (20, 50.4 g/L) at acidities (1–4 M HNO3), and simulated Advanced Heavy Water Reactor (AHWR) spent fuel dissolver solution. Experiments were also carried out to recover uranium from assorted organic waste solutions. Dispersion number of DHOA was calculated prior to counter-current runs and compared with different solvents proposed for back-end of fuel cycle. These runs demonstrated that DHOA is a promising alternative of TBP for various hydrometallurgical operations dealing with the recovery of uranium from solutions of various origins including spent fuel dissolver solutions. [ABSTRACT FROM AUTHOR]

An Advanced Heavy Water Reactor (AHWR) has been specifically designed to exploit Th/233U as fuels. The reprocessing is focused mainly on the recovery of233U and Pu from the spent fuels leaving bulk of Th (∼100 g/L) in the High Level Waste (HLW) solutions. No systematic attempts have been made so far to identify suitable solvents for the recovery of thorium from high level waste (HLW) solutions generated after AHWR spent fuel reprocessing. Tri-n-butyl phosphate (TBP), though the work horse for nuclear fuel reprocessing as an extractant, suffers from the serious limitation of third-phase formation during the extraction of macro concentrations of thorium. Two straight chain dialkyl amides such asN,N-dihexyl octanamide (DHOA), andN,N-dihexyl decanamide (DHDA) as well as TBP were evaluated for the recovery of the thorium from AHWR-HLW solutions. Attempts were made to identify suitable solvent (extractant + diluent) system and optimize the conditions for the recovery of thorium from HLW solutions. Selectivity of the solvents was examined for thorium extraction over fission products/structural materials under AHWR raffinate solutions. Counter-current centrifugal contactor runs were also carried out on simulated waste solutions to validate the optimized conditions for the recovery of thorium from the simulated AHWR waste solutions. [ABSTRACT FROM PUBLISHER]

Extraction behavior of neptunium has been compared for tri-n-butyl phosphate (TBP) and N,N-dihexyl octanamide (DHOA) extractants as a function of nitric acid concentration (0.5 − 6 M HNO3), uranium loading (50 and 300 g/L relevant to Pu rich fast reactor and Pressurized Heavy Water Reactor, PHWR spent fuels, respectively), and in the presence of oxidizing and reducing agents. These studies suggest the possibility of co-recovery of U(VI), Pu(IV) and Np(IV) from spent fuel dissolver solutions (of Pu rich fuels) employing DHOA as extractant. [ABSTRACT FROM PUBLISHER]

Straight chain N,N-dihexyloctanamide (DHOA) has been identified as a promising alternate extractant to tributyl phosphate (TBP) for the reprocessing of uranium based spent fuels. The present work compares extraction behavior of technetium using DHOA and TBP solutions in n-dodecane, under varying experimental conditions such as acidity (0.5-6 M HNO3); extractant concentration (1.1 and 1.5 M), and uranium loading (50 g/L, relevant for Pu rich spent fuel feed solutions). The effect of acetohydroxamic acid concentration on U, Pu, Np, and Tc extraction behavior has also been investigated. Pu(IV)-AHA interaction and its influence on extraction using TBP and DHOA extractants has been studied spectrophotometrically. The experimental data suggest that 1.1 M DHOA is better than 1.1 M TBP with respect to co-extraction of Tc and U, and U decontamination with respect to Np/Pu. [ABSTRACT FROM AUTHOR]