S. Panja

Transport of Thorium(IV) Across a Supported Liquid Membrane Containing N,N,N′,N′-Tetraoctyl-3-oxapentanediamide (TODGA) as the Extractant

The transport behavior of Th4+ was investigated from a feed containing 3.0 M HNO3 into a receiver phase containing 0.1 M oxalic acid across a PTFE flat sheet supported liquid membrane (SLM) which contained TODGA (N,N,N′,N′-Tetraoctyl-3-oxapentanediamide) in n-dodecane as the extractant. Effects of the nature of the strippant, extractant concentration, Th concentration in the feed, and feed acidity on the transport rates were investigated. The transport behavior apparently depended on the rate of extraction of the metal ion at the feed-membrane interface and was not diffusion controlled. Influence of Th concentration on flux was also investigated. Transport mechanism was elucidated and the diffusion coefficient was calculated to be 2.13 × 10−7 cm2/s. Solvent extraction studies at varying feed acidity and TODGA concentration were also carried out.

Controlled Pertraction of Plutonium(III) Under Reducing Conditions From Acidic Feeds Using TODGA as the Carrier Extractant

Transport of Pu3+ across PTFE flat sheet supported liquid membranes (SLM) using N,N,N′,N′-tetraoctyl-3-oxapentanediamide (TODGA) in n-dodecane as the carrier extractant was investigated. The feed solution was usually 3.0 M HNO3 while the receiver phase contained 0.1 M HNO3. In order to model the transport data, the two-phase liquid-liquid extraction experiments are also reported. A series of reducing agents viz. hydrazinium nitrate (HN), phenyl hydrazine (PH), hydroxyl ammonium nitrate (HAN), and ascorbic acid (AA) were evaluated to maintain plutonium in the +3 state in the feed, out of which phenyl hydrazine was found to be the most effective. Effectiveness of various strippants for quantitative and rapid transport of Pu was also evaluated. Additionally, the effects of feed acidity, carrier concentration, membrane thickness, effect of Pu carrier concentration, etc. were investigated. Membrane diffusion co-efficients were calculated and found to be 1.47 × 10−6 cm2/s for Pu(III). The stability of the SLM, investigated as a function of contact time, was found to be reasonably good even after 20 days of continuous operation.

Carrier Facilitated Transport of Europium(III) Across Supported Liquid Membranes Containing N,N,N′,N′-tetra-2-ethylhexyl-3-oxapentane-diamide (T2EHDGA) as the Extractant

Studies on the solvent extraction and pertraction behavior of europium(III) was carried out from acidic feed solutions using N,N,N′,N′-tetra-2-ethylhexyl-3-oxapentane-diamide (T2EHDGA) in n-dodecane as the solvent. The nature of the extracted species from the solvent extraction studies conformed to Eu(NO3)3 · 3T2EHDGA which is in variance with the analogous Eu(III) – TODGA (linear homolog of T2EHDGA) extraction system. The transport behavior of Eu(III) was investigated from a feed containing 3.0 M HNO3 into a receiver phase containing 0.01 M HNO3 across a PTFE flat sheet supported liquid membrane (SLM) containing 0.2 M T2EHDGA in n-dodecane as the carrier solvent and 30% iso-decanol as the phase modifier. Effects of feed acidity, carrier extractant concentration, membrane pore size, and Eu concentration in the feed on the transport rates of Eu(III) were also investigated. Membrane diffusion coefficient (D o) for the pertracted species was calculated using the Wilke-Chang equation as 4.25 × 10−6 cm2 · s−1. The influence of Eu concentration on the flux was also investigated. The role of temperature on the transport rates was investigated and the thermodynamic parameters were calculated.

Studies on the Development of a Flow-Sheet for AHWR Spent Fuel Reprocessing Using TBP

The present paper describes the results of solvent extraction studies carried out in batch mode to collect data on distribution of uranium, plutonium, and thorium using 5% TBP in n-dodecane. Extraction studies are carried out from feed solutions having bulk thorium containing aluminum and fluoride ions in ∼3.00–4.00 M nitric acid at concentration levels anticipated in feed solutions during Advanced Heavy Water Reactor (AHWR) spent fuel reprocessing. Studies are carried out under varied experimental conditions. Parameters such as organic to aqueous phase ratio during extraction, concentration of nitric acid for scrubbing co-extracted thorium from loaded organic phase etc., are studied in detail. Hydroxylamine nitrate is selected for reductive stripping of plutonium in preliminary studies. Reagent mixture containing 0.30 M HAN + 0.60 M HNO3 and 0.20 M N2H4 is found to be optimum for plutonium partitioning. This paper also describes the extraction and stripping of uranium and plutonium in co-current mode. The extraction behavior of relevant fission products is studied from a simulated feed solution. A preliminary study on a few commercially available reducing agents is also included. These data are useful in developing a flow-scheme for the recovery of uranium and plutonium from spent fuel originating from AHWR.

Solvent Extraction of Pu(IV) Using TBP: A Comparative Study of n-dodecane and a Room Temperature Ionic Liquid

The extraction of Pu(IV) from nitric acid medium using TBP has been compared with two diluents using n-dodecane and a RTIL, 1-butyl-3-methylimidazolium bis(trifluoromethanesulphonyl)imide (C4mimTf2N). Kinetics of extraction was slower in RTIL compared to n-dodecane. The mechanism of extraction was found to be cation exchange at lower acidity which changes to the formation of a neutral complex at higher acidity greater than 1M HNO3 for RTIL. Back extraction of Pu from the loaded organic phase was equally effected using 0.1M oxalic acid and a mixture of guanidine carbonate and EDTA as strippant from both the extraction systems under study. Radiation stability and reusability studies showed promise for future application.

Facilitated transport of uranium(VI) across supported liquid membranes (SLM) containing T2EHDGA: Part II. Nature of feed, pore size and temperature on pertraction rates

ABSTRACT Facilitated transport of uranyl ion was investigated across PTFE supported liquid membranes using T2EHDGA in n-dodecane as the carrier extractant containing 30% iso-decanol as the phase modifier. The feed was usually 3 M HNO3 while the strip phase constituted pH 2 solution. About 11 h were required for quantitative transport of U(VI) from a feed of 3 M HNO3 using 0.2 M T2EHDGA in n-dodecane containing 30% iso-decanol as the carrier extractant. Transport studies carried out using various mineral acid as feed indicated the transport trend as HNO3 >> HCl ∼ HClO4 which did not change significantly when the respective sodium salts were used in the feed. Effect of various parameters such as membrane pore size, U concentration in the feed and temperature, etc. on the transport rate was also investigated. Comparisons were made with the transport behaviour of plutonyl ion and TODGA system.

Validation of the flow-sheet proposed for reprocessing of AHWR spent fuel: counter-current studies using TBP

ABSTRACT A solvent extraction based flow-sheet, for reprocessing of spent fuel arising from Advanced Heavy Water Reactor (AHWR) for separation of uranium, plutonium and thorium using 5% TBP in n-dodecane, has been tested using laboratory scale mixer-settlers. Simulated feed solution containing thorium, uranium and plutonium in the concentration levels expected in feed solutions of AHWR spent fuel reprocessing is prepared using plutonium, natural uranium and thorium nitrate. Quantitative extraction of uranium and plutonium is achieved under experimental conditions leaving bulk of thorium in the raffinate. Co-extracted thorium from the organic phase is scrubbed using 3.00 M HNO3. Separation of plutonium from uranium is achieved by chemical reduction employing a mixture of hydroxylamine nitrate and hydrazine nitrate in nitric acid. Results show quantitative partitioning. Uranium from the plutonium lean organic phase is stripped using 0.01 M HNO3. The results clearly established the validity of the proposed f...

BenzoDODA grafted polymeric resin—Plutonium selective solid sorbent

A new ligand grafted polymeric resin (BenzoDODA SDVB) was synthesized by covalently attaching plutonium selective ligand (BenzoDODA) on to styrene divinyl benzene (SDVB) polymer matrix. BenzoDODA SDVB resin was evaluated for separation and recovery of plutonium(IV) from nitric acid medium. Sorption of Pu(IV) was found to decrease with the increase in nitric acid concentration, with very small sorption above 7.0M HNO3. Sorption kinetics was fast enough to achieve the equilibrium within 60min of contact where the kinetic data fitted well to pseudo-second-order model. Sorption isotherm data fitted well to Langmuir model suggesting chemical interaction between the BenzoDODA moiety and plutonium(IV) ions. Sorption studies with some of representative radionuclides of high level waste showed that BenzoDODA SDVB is selective and therefore could be a promising solid sorbent for separation and recovery of plutonium. Further, the theoretical calculations done on BenzoDODA SDVB resin suggested Pu(NO3)4·BenzoDODA (1:1) sorbed complex conformed to generally observed square antiprism geometry of the plutonium complexes, with contributions from oxygen atoms of four nitrate ions as well as from four oxygen atoms present in BenzoDODA (two phenolic ether oxygen atoms and two carbonyl oxygen atoms of amidic moiety).

Liquid – Liquid Extraction and Pertraction of Eu(III) from Nitric Acid Medium Using Several Substituted Diglycolamide Extractants

Solvent extraction and supported liquid membrane (SLM) transport properties of Eu(III) from nitric acid feed conditions were investigated using several substituted diglycolamide (DGA) extractants such as N,N,N′N′-tetra-n-octyl diglycolamide (TODGA), N,N,N′N′-tetra(2-ethylhexyl) diglycolamide (T2EHDGA), N,N,N′N′-tetra-n-hexyl diglycolamide (THDGA), N,N,N′N′-tetra-n-pentyl diglycolamide (TPDGA), and N,N,N′N′-tetra-n-decyl diglycolamide (TDDGA). Effects of feed acidity and phase modifier composition on Eu(III) extraction were investigated using the DGAs and the nature of extracted species were ascertained by slope analysis method. The Eu(III) distribution ratio (DEu) values were found to decrease in the presence of iso-decanol. In general, the DEu values decreased with increased alkyl chain length of the DGA. The extracted species contained only 2 extractant molecules when TPDGA and TDDGA were used while for TODGA about four extractant molecules were found to be present in the extracted species. The supported liquid membrane transport of Eu(III) was studied under varying experimental conditions using the five DGA extractants. Transport studies using 0.1 M DGA as the extractant suggested the trend as TDDGA > TODGA > T2EHDGA ∼ THDGA which significantly changed to TPDGA > THDGA > TODGA > TDDGA > T2EHDGA in the presence of 30% iso-decanol as the phase modifier. The permeability coefficient (P) values were also determined with membranes of varying pore sizes.

Selective separation of Cs(I) extraction from actual high level waste using a solvent containing calix [4]-bis 2,3-naphtho-crown-6

ABSTRACT Solvent extraction studies were carried out on the selective separation of radio-cesium from actual high level waste (HLW) using calix [4]-bis-2,3-naphtho-crown-6 (CNC) in nitrobenzene - toluene mixture. The separation studies were carried out in two stages. In the first stage, a 100 times diluted HLW was used and the purity was ascertained by gamma ray spectrometry using a HPGe detector. Quantitative separation of radio-cesium (monitored by Cs-137 gamma peak at 661 keV) was observed with almost no contamination from any of the other fission products present. In the second stage, actual HLW was used for the Cs separation study. Four stages of extraction and two stages of stripping using distilled water gave >99% recovery. About 10 mCi Cs was recovered by this method. Reusability of the solvent was also carried out and though there was no loss in selectivity, slight decrease in the extraction efficiency was observed after the reagent was allowed to be in contact with the HLW for 10 d.