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Abstract  

Imidazolium nitrate anchored on poly(styrene-divinylbenzene) co-polymer, Im-NO3, has been synthesized and evaluated for plutonium purification. The results are compared with those obtained using Dowex 1 × 4 anion exchange resin. The distribution coefficient (Kd) of Pu(IV) increased with increase in concentration of nitric acid, reached a maximum at 8 M, followed by decrease in Kd values. Rapid ion exchange of Pu(IV) followed by the establishment of equilibrium occurred within 100 min of equilibration and the data was fitted in to first order rate equation. Variation of distribution coefficient of Pu(IV) as a function of exchange capacity and nitrate ion concentration suggest the involvement of anion exchange mechanism is responsible for extraction. The apparent ion exchange capacity was 310 mg/g at 8 M nitric acid. The performance of the Im-NO3 under dynamic condition was assessed by column breakthrough experiments. Radiolytic degradation of Im-NO3 resin in presence and absence of nitric acid (8 M) was studied and the results are reported in this paper.

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Abstract  

High-level liquid waste from fast reactor fuel reprocessing stream contains significant quantities of lanthanides and trivalent minor actinides. The lanthanides and minor actinides (MA) have been separated from the fast reactor high-level liquid waste (FR-HLLW) using TRUEX solvent, which is a mixture of 0.2 M octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO)-1.2 M tri-n-butylphosphate (TBP) in n-dodecane. A new stripping composition, 0.1 M HNO3 and 0.1 M citric acid (CA), has been employed for back extraction of them from the TRUEX solvent. In order to separate lanthanides from actinides present in the strip solution, the extraction behavior of 241Am(III) and (152+154)Eu(III) from CA–HNO3 medium by a solution of bis-2-ethylhexylphosphoric acid (HDEHP) in n-dodecane has been studied. Separation factors (SF = D Eu/D Am) has been reported as a function of various parameters such as pH, concentrations of HDEHP, diethylenetriamine-N,N,N′,N′′,N′′′-pentaaceticacid (DTPA), 1-octanol and TBP in this paper.

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Summary  

A systematic study on the extraction of U(VI) from nitric acid medium by tri-n-butylphosphate (TBP) dissolved in a non-traditional diluent namely 1-butyl-3-methylimidazolium hexafluorophosphate (bmimPF6) ionic liquid (IL) is reported. The results are compared with those obtained using TBP/n-dodecane (DD). The distribution ratio for the extraction of U(VI) from nitric acid by 1.1M TBP/bmimPF6 increases with increasing nitric acid concentration. The U(VI) distribution ratios are comparable in the nitric acid concentration range of 0.01M to 4M, to the ratios measured using 1.1M TBP/DD. In contrast to the extraction behavior of TBP/DD, the D values continued to increase with the increase in the concentration of nitric acid above 4.0M. The stoichiometry of uranyl solvate extracted by 1.1M TBP/IL is similar to that of TBP/DD system, wherein two molecules of TBP are associated with one molecule of uranyl nitrate in the organic phase. Ionic liquid alone also extracts uranium from nitric acid, albeit to a small extent. The exothermic enthalpy accompanying the extraction of U(VI) in TBP/bmimPF6 decreases with increasing nitric acid and with TBP concentrations.

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Cereal Research Communications
Authors: M. Sivasamy, M. Aparna, J. Kumar, P. Jayaprakash, V.K. Vikas, P. John, R. Nisha, K. Srinivasan, J. Radhamani, S.R. Jacob, S. Kumar, Satyaprakash, K. Sivan, E. Punniakotti, R.K. Tyagi, and K.C. Bansal
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