Poonam Jagasia

Separation of Carrier-Free 90Y from 90Sr by SLM Technique Using D2EHPA in N-Dodecane as Carrier

Solvent extraction studies on Sr2+ and Y3+ are carried out from varying concentrations (0.01–6.0 M) of nitric acid using di-(2-ethylhexyl)phosphoric acid (D2EHPA) as extractant. Extraction of yttrium is observed to be higher than that of strontium at all the acidities and is found to increase substantially with decreasing concentration of nitric acid. Practically negligible extraction (D < 10−3) of Sr2+ is observed from feed solutions containing nitric acid in the range of 1.0 to 4.0 M. These solvent extraction data are used to optimize the transport of 90Sr and 90Y across the supported liquid membrane (SLM) individually as well as from their mixture (due to insitue growth) under different experimental conditions. Selective separation of 90Y (>90%) from 90Sr is obtained in 6 h, when the concentration of nitric acid in feed is kept at 1.0 M and that of receiving phase is maintained at 4.0 M. 20% D2EHPA in n-dodecane is found to be the optimum carrier concentration for the efficient transport of 90Y in SLM mode. Under these conditions transport of strontium is found to be negligible. Radiochemical purity of the product 90Y is checked by following its decay as well as by extraction paper chromatography. The contamination of 90Sr in 90Y product is found to be < 0.001%. Based on the experimental results, a single stage SLM system for the generation of carrier-free 90Y from 90Sr source is described. The system is amenable for automation and scale up.

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.

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...

Evaluation of Novel Solvent Systems Containing Calix-crown-6 Ligands in A Fluorinated Solvent for Cesium Extraction from Nitric Acidic Feeds

Several commercial calix-crown-6 ligands such as: calix[4]arene-bis-crown-6 (CC), calix[4]arene-bis-benzo-crown-6 (CBC), calix[4]arene-bis-naphtho-crown-6 (CNC), and bis-(octyloxy)calix[4]arene-mono-crown-6 (CMC) have been evaluated for their Cs uptake ability from nitric acid feed solutions in a novel solvent system. The calix-crown-6 ligands were made soluble in phenyl trifluoromethyl sulphone (PTMS) and 1 × 10−3 M solutions were used for metal ion extraction from nitric acid feed solutions. The extraction efficiency of the calix-crown-6 ligands was found out to be: CBC > CNC > CMC ˜ CC, which was attributed to trends of their partition coefficients. CNC was considered to be the most suitable extractant in view of its better organophilicity. The effects of concentrations of nitric acid and Cs carrier in feed were also investigated. Studies carried out using variation of ligand concentration yielded an extracted species of the type [CsL+][NO3−]org. Stripping of Cs(I) from the loaded organic phase was found to be possible when the organic phase contained minute quantities of Alamine 336. Co-current extraction and stripping studies were also carried out using CNC as the extractant.

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.

Separation and purification of 106Ru from effluent streams of ion-exchange cycle used for Pu purification in PUREX process

AbstractThis paper describes the separation and purification of 106Ru from ion-exchange effluent streams of PUREX process using combinations of solvent extraction, extraction chromatography, and ion-exchange. The effluent stream from ion-exchange cycle contains ∼100 mCi/L 106Ru activity along with contaminants like 95Nb, uranium, and plutonium at an acidity of ∼7.00 M HNO3. In first step, the feed solution is contacted with 30% Tri-n-butyl phosphate in n-dodecane which not only lowers the acidity of the feed solution but also reduces the Pu, U, and Nb concentration significantly. The raffinate from this step is subjected to extraction chromatography wherein uranium and plutonium are completely sorbed on a column containing 2-ethylhexyl 2-ethylhexyl phosphonic acid (KSM-17) sorbed on an inert polymeric support XAD-4. In these steps, Ru activity is found to retain in the aqueous raffinate/effluent phase without any loss. The effluent from the second step is neutralized to an acidity of ∼0.50 M using NaOH so...