P. K. Dey

Pseudo-Emulsion Based Hollow Fiber Strip Dispersion Technique (PEHFSD): Optimization, Modelling and Application of PEHFSD for Recovery of U(VI) from Process Effluent

Abstract Pseudo emulsion based hollow fiber strip dispersion technique (PEHFSD) is the first of its kind ever explored in radioactive environment for the extraction of uranium from acidic process streams. Permeation of U(VI) was investigated as a function of various experimental variables such as hydrodynamic conditions (flow rates of pseudo-emulsion and feed phase), concentration of U(VI) in the feed phase, concentration of tri-n-butylphosphate (TBP), HNO3 concentration in feed phase, O/A ratio and 0.01 M HNO3 as stripping agent in pseudo-emulsion phase. The mass transfer coefficient was calculated from the experimental results and a model has been presented for determining mass transfer characteristics. PEHFSD has been demonstrated for separation/recovery of uranium from oxalate supernatant waste generated during plutonium precipitation by oxalic acid. PEHFSD and HFSLM (hollow fiber supported liquid membrane) performance has been compared in order to analyze the efficiency of the technique.

Studies on the Development of a Two Stage SLM System for the Separation of Carrier‐free 90Y using KSM‐17 and CMPO as Carriers

Abstract Solvent extraction studies of Y3+ and Sr2+ with 2‐ethylhexyl 2‐ethylhexyl phosphonic acid (KSM‐17) and octyl(phenyl)‐N,N‐diisobutylcarbamoylmethylphosphine oxide (CMPO) are carried out from aqueous media containing a wide range of nitric acid and other potential reagents to arrive at the operating conditions for the selective transport of 90Y using supported liquid membrane (SLM) containing these reagents as carriers. Since the transport data of 90Y using single cell SLM with KSM‐17 was available from our earlier experiments, single cell transport studies with CMPO carrier are only carried out to optimize the strippant phase. Transport studies with pure 90Y is carried out using a transport cell with two SLMs one with KSM‐17 and the other CMPO carriers to optimize the transport parameters. Based on these data the development of a two stage SLM system for the generation of carrier free 90Y from 90Sr source is described. The procedure described is amenable for automation and scale up.

Application of Hollow Fiber Contactor in Nondispersive Solvent Extraction of Pu(IV) by TBP

Abstract Data on dispersion‐free solvent extraction (DFSX) of Pu(IV) from acidic nitrate media using microporous hydrophobic polypropylene hollow fiber membrane contactor with tri‐n‐butylphosphate (TBP) employing as an extractant are being presented. The DFSX operation was carried out with various concentration of TBP in n‐dodecane by passing acidic feed containing Pu (IV) through the tube side at the flow rate of 5.83 cm3 s−1 and organic extractant through the shell side at the flow rate of 1.53 cm3s−1. Extraction studies were performed under different hydrodynamic conditions and the overall mass‐transfer coefficient was evaluated with countercurrent flow condition. It was possible to extract and concentrate Pu(IV) from aqueous phase by employing this technique. For back extracting the Pu, uranous nitrate and hydroxylamine hydrochloride solution as strippants were examined, which flowed through the tube side (flow rate: 6.11 cm3 s−1) and the loaded organic was passed through the shell side with the flow rate of 1.66 cm3s−1. Results revealed that ∼80% of Pu(IV) from oxalate supernatant waste could be extracted by utilizing this technique.

Separation of Uranium and Plutonium from Aqueous Acidic Wastes Using a Hollow Fiber Supported Liquid Membrane

Abstract The use of microporous hydrophobic polypropylene, as a hollow fiber supported liquid membrane (HFSLM) was considered for removal of actinides such as uranium (U) and plutonium (Pu) from nuclear process effluents. The 1.09 M extractant, tri‐n‐butyl phosphate (TBP) diluted with n‐dodecane was used as carrier. The study includes the hydrodynamic and chemical parameters. Modeling was performed using chemical parameters and rate controlling steps were identified. It was possible to remove U(VI) and Pu(IV) from process effluent more than 99% in presence of fission products. The optimum effective feed linear flow velocity was found to be 0.88 cm sec−1. The stripping reagent 0.1 M hydroxylamine hydrochloride (NH2OH · HCl) in 0.5 M HNO3 was used. The permeation of metal ions increased with increasing effective surface area and model for higher concentration of metal ion was able to describe the transport mechanism of U(VI).

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.

Radiochemical Separation and Purification of 144Ce from Purex High‐Level Waste

ABSTRACT A solvent extraction technique has been developed for the radiochemical separation and purification of 144Ce from Purex high-level waste (HLW). The feed solution initially was in contact with granulated ammonium molybdophosphate for the selective removal of cesium to reduce the gamma dose during the subsequent process. In the next step, uranium and plutonium were removed quantitatively by using 20% 2-ethylhexyl hydrogen 2-ethylhexyl phosphonate (KSM-17) in dodecane. The trivalent cerium present in the cesium-, uranium-, and plutonium-depleted raffinate was oxidized to its tetravalent state by using K2Cr2O7 and was extracted into KSM-17. The loaded cerium from the organic phase was stripped with 0.5 M nitric acid. Final decontamination from trace impurities such as 106Ru and 125Sb could be achieved by an additional cerium-extraction step with KSM-17 at pH 2, followed by its stripping with 0.5 M HNO3. About 10 mCi of cerium was separated from Purex HLW by using this technique. This method is amenable for automation and scale up.

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.

Some Parametric Studies on Separation of Palladium from Perchloric Acid Medium by Radiolytic Reduction

Abstract The radiolytic reduction of palladium ions to palladium metal in perchloric acid solutions has been studied with the aim of separating palladium from aqueous acidic waste. Fraction of Pd separated out as precipitate has been studied as a function of initial Pd concentration and strength of HClO4. Addition of t‐butanol to the system has been found to cause substantial enhancement in the amount of Pd precipitated as compared to that in its absence. At a given absorbed dose, the extent of Pd separated is found to increase with the concentration of HClO4 in presence of t‐butanol. However, the converse is true for the radiolysis in absence of t‐butanol. The decrease in the extent of reduction of Pd(II) to Pd(0) has been found to be due to increasing formation of chloride ions that tend to form reduction resistant cholorocomplexes of Pd. This is supported by the red‐shifting in the absorption bands of Pd(ClO4)2 observed for the spectra of gamma radiolysed solutions of Pd(II) at higher HClO4 concentrations. External addition of chloride ions to aqueous Pd(II)/HClO4 system even in presence of t‐butanol has been found to cause substantial inhibition to the radiolytic reduction of Pd(II) owing to formation of reduction‐resistant complexes. Correspondingly, the addition of nitrate ions to Pd(II)/HClO4/t‐butanol system showed inhibition effect at much greater stoichiometric amount of nitrate.