P. V. Achuthan

Complexation studies with 90Y from a novel 90Sr-90Y generator

Some features of a novel 90Sr-90Y generator which employs supported liquid membrane (SLM) to separate carrier-free 90Y from 90Sr present in the high level waste of the spent fuel of reactor are described. After ascertaining the purity of 90Y particularly with respect to 90Sr breakthrough, its complexation was studied with a few oxo/aza donor ligands, such as DTPA, EDTMP, DOTA, TETA and a cyclic phosphonate, CTMP. These studies were primarily carried out to adjudge the quality of the 90Y obtained from a novel 90Sr-90Y generator and ascertain its usability for labelling biomolecules such as antibodies and peptides. The DOTA complexes are most stable at 37 C in human serum; they appear to be ideal bifunctional chelating agent for use in radioimmunotherapy with 90Y.

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.

Separation of Carrier-Free 90Y from High Level Waste by Extraction Chromatographic Technique Using 2-Ethylhexyl-2-ethylhexyl Phosphonic Acid (KSM-17)

An extraction chromatographic technique has been developed for the separation of carrier-free 90Y from the 90Sr present in the high level waste (HLW) of the Purex process. When a Purex HLW solution in 2–3 M HNO3 is passed through a CMPO-Chromosorb-102 (CAC) column, all the trivalent, tetravalent, and hexavalent ions are sorbed. The effluent from this experiment, after adjusting the pH to 2 with NaOH, was passed through a 2-ethylhexyl-2-ethylhexyl phosphonic acid (KSM-17)-Chromosorb-102 (KSMC) extraction chromatographic column where only 90Y was sorbed. All the other ions (90Sr, 137Cs, 125Sb, 106Ru, 106Rh, etc.) were washed off with dilute HNO3 (pH 2), and carrier-free 90Y was eluted with 0.5 M HNO3. This technique can yield 90Y in mCi levels in pure form for medical applications. The 90Sr can be used repeatedly after allowing for 90Y buildup.

Effect of Diluents on the Extraction of Sr2+ from HNO3 Solutions with Dicyclohexano‐18‐crown‐6

Abstract Extraction of Sr2+ has been carried out with dicyclohexano‐18‐crown‐6 (DC18C6) in benzene from varying HNO3 concentrations, and the nature of species extracted has been established. The equilibrium constant values for the formation of Sr(NO3)2·DC18C6 have been evaluated, taking into account the aqueous complexing of Sr2+ with NO3 − and with aqueous‐soluble DC18C6, and also the uptake of HNO3 by DC18C6 in the organic phase. Effect of diluents apart from benzene, such as chlorobenzene, dichloroethane, chloroform, 1‐octanol, carbon‐tetrachloride, and nitrobenzene, and various mixtures of benzene and nitrobenzene on the extraction of Sr2+ with DC18C6 has also been studied. On the practical side, the extraction of Sr2+ has been examined with DC18C6 in solvent mixtures of tri‐n‐butyl phosphate (TBP), dodecane, and 1‐octanol. This solvent mixture, at least partially, avoids the disadvantages envisaged while using 1‐octanol alone for the separation and recovery of Sr2+ from minor actinide depleted TRUEX process raffinate.

EXTRACTION CHROMATOGRAPHIC SEPARATION OF PROMETHIUM FROM HIGH ACTIVE WASTE SOLUTIONS OF PUREX ORIGIN

ABSTRACT An extraction chromatographic procedure for the separation of 147Pm from High Active Waste solutions of Purex process has been developed. Octy1(pheny1)-N,N-diisobutylcarbamoylmethy1phosphine oxide(CMPO) and 2-ethylhexy1-2-ethylhexylphosphonic acid (KSM-17), both sorbed separately on an inert support(chromosorb-102) have been sequentially employed for this purpose. In the CMPO column, the rare earths and the trivalent actinides are sorbed together with uranium, plutonium and traces of few other fission products. The elution of this column with 0.04 M HNO3 gives an eluate containing trivalent actinides and lanthanides. This solution, after adjusting the pH to 2.0, is used as feed for the second extraction chromatographic column based on KSM-17. All the trivalent metal ions are sorbed on the column leaving the trace impurities in the effluent. Fractional elution of the metal ions from this column is carried out with nitric acid of varying concentrations, At 0.09 M HNO3, the pure beta emitting fracti...

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.