Search Results

You are looking at 31 - 40 of 45 items for

  • Author or Editor: V. Manchanda x
  • Refine by Access: All Content x
Clear All Modify Search

Abstract  

Release of long-lived radioactivity to the aquatic bodies from various nuclear fuel cycle related operations is of great environmental concern in view of their possible migration into biosphere. This migration is significantly influenced by various factors such as pH, complexing ions present in aquatic environment and sorption of species involving radionuclides on the sediments around the water bodies. 241/243Am are two major radionuclides which can contribute a great deal to radioactivity for several thousand years. In the present study, 241Am sorption on natural sediment collected from site near a nuclear installation in India, has been investigated under the varying conditions of pH (3–10) and ionic strength [I = 0.01–1 M (NaClO4)]. The sorption of Am increased with pH of the aqueous medium [10% (pH 2) to ~100% (pH 10)], which was explained in terms of the increased negative surface charge on the sediment particles. There was marginal variation in Am(III) sorption with increased ionic strength (within error limits) of the aqueous medium suggesting inner-sphere complexation/sorption process. Sediment was characterized for its elemental composition and structural phases using Energy Dispersive X-Ray (SEM-EDX) and X-Ray Diffraction (XRD) techniques. Zeta-potential measurement at I = 0.1 M (NaClO4) suggested that Point of Zero Charge (pHPZC) was ~2, indicating the presence of silica as major component in the sediment. Kurabtov plot using sorption data as a function of pH at fixed I = 0.1 M (NaClO4) indicated the presence of multiple Am(III) species present on the surface. Potentiometric titration of the suspension indicated the presence of mineral oxide like behavior and assuming a generic nature (≡XOH) for all types of surface sites, protonation–deprotonation constants and total number of sites have been obtained. The sorption data has been modeled using 2-pK Diffuse Double Layer Surface Complexation Model (DDL-SCM). ≡XOAm2+ has been identified as the main species responsible for the sorption profile.

Restricted access

Abstract  

The primary purpose of this study was to understand the alpha radiolytic degradation behavior of N,N-dihexyl octanamide (DHOA) vis a vis tributyl phosphate (TBP) solutions in n-dodecane under plutonium loading conditions. These studies were carried out as a function of dose on different Pu loaded samples (containing 0.002-10 g/L Pu) from 4 M HNO3 medium. These Pu loaded solutions were evaluated for stripping behavior by contacting with 0.5 M NH2OH at 0.5 M HNO3 solutions. Organic phase analysis was carried out by gas chromatography (GC) and by visible spectrophotometry. These studies clearly indicated that Pu stripping becomes difficult with increased dose in the case of TBP system. On the other hand, no such problem was observed in DHOA system during stripping of plutonium, thereby indicating that DHOA is a promising candidate for the reprocessing of high burn up Pu rich spent fuels.

Restricted access

Abstract  

Extraction behavior of 1 × 10−2–0.1 M U(VI) from aqueous phases containing 0.86 M Th(IV) at 4 M HNO3 in 1.1 M tributyl phosphate (TBP) and 1.1 M N,N-dihexyl octanamide (DHOA) solutions in different diluents viz. n-dodecane, 10% 1-octanol + n-dodecane, and decahydronaphthalene (decalin) was studied. Third-phase formation was observed in both the extractants using n-dodecane as diluent. There was a gradual decrease in Th(IV) concentration in the third-phase (heavy organic phase, HOP) with increased aqueous U(VI) concentration [0.71 M (no U(VI))–0.61 M (0.1 M U(VI)) for TBP; 0.27 M (no U(VI))–0.22 M (0.1 M U(VI)) for DHOA]. The HOP volume in case of DHOA was ~2.2 times of that of TBP. Uranium concentration in HOP increased with its initial concentration in the aqueous phase [from 1.8 × 10−2 M (0.01 M U(VI))–0.162 M (0.1 M U(VI)) for TBP; from 1.4 × 10−2 M (0.01 M U(VI))–0.14 M (0.1 M U(VI)) for DHOA] suggesting that Th(IV) was being replaced by U(VI). An empirical correlation was developed for predicting the concentrations of uranium and thorium in HOP for both the extractants. No third-phase appeared during the extraction of uranium and thorium from the aqueous phases employing 10% 1-octanol + n-dodecane, or decalin as diluents, and therefore, were better choices as diluent for alleviating the third-phase formation during the reprocessing of spent thorium based fuels, and for the recovery of thorium from high-level waste solutions.

Restricted access

Abstract  

The estimation of low level alpha activity is difficult in waste samples containing large concentration of salts and beta–gamma activity. In the present study, the feasibility of gross alpha-activity measurement for simulated high level waste (SHLW) in solution medium by alpha-track registration technique has been attempted. The results showed that it is possible to use this technique for gross alpha-activity estimation of ~200 Bq/mL in solution medium with a precision and accuracy of ~30%. The importance of measuring 200 Bq/mL alpha activity in SHLW solutions is that this value corresponds to about 4,000 Bq/g activity in the solid medium which is the safe disposable limit. The advantage of this method over other methods is that it is not sensitive to beta–gamma emitters and salts and is very simple and inexpensive.

Restricted access

Abstract  

The extraction of U(VI) from sulphate medium with 2-ethylhexyl phosphonic acid-mono-2-ethylhexyl ester (PC88A, H2A2 in dimeric form) in n-dodecane has been investigated under varying concentrations of sulphuric acid and uranium. Slope analysis of uranium (VI) distribution data as a function of PC88A concentration suggests the formation of monomeric species, viz. UO2(HA2)2. This observation was further supported by the mathematical expression obtained during non-linear least square regression analysis of U(VI) distribution data correlating the percentage extraction (%E) and the acidity (H i). A mathematical model correlating the experimental distribution ratio values of U(VI) (D U) with initial acidity (H i) and initial uranium concentrations (C i) was developed:
\documentclass{aastex} \usepackage{amsbsy} \usepackage{amsfonts} \usepackage{amssymb} \usepackage{bm} \usepackage{mathrsfs} \usepackage{pifont} \usepackage{stmaryrd} \usepackage{textcomp} \usepackage{upgreek} \usepackage{portland,xspace} \usepackage{amsmath,amsxtra} \pagestyle{empty} \DeclareMathSizes{10}{9}{7}{6} \begin{document} $$D_{\text{U}} = 12.98( \pm 0.90)/\left\{ {C_{\text{i}}^{ - 0.75( \pm 0.05)} \times \left[ {H_{\text{i}} } \right]^{2} } \right\}$$ \end{document}
. This expression can be used to predict the concentration of uranium in organic as well as in aqueous phase at any C i and H i. The extraction data were used to calculate the conditional extraction constant (K ex) values at different acidities (2–7 M H+), uranium (0.02–0.1 M) and PC88A (0.2–0.6 M) concentrations. These studies were also extended for the extraction of U(VI) using synergistic mixtures of PC88A and TOPO from sulphate medium.
Restricted access

Abstract  

The potentiometric determination of uranium is widely carried out in phosphoric acid medium to suppress the interferences of plutonium by complexation. Owing to the complexity of the recycling plutonium from the phosphate based waste involving manifold stages of separation, a method has been proposed in the present paper which does not use phosphoric acid. Uranium and plutonium are reduced to U/IV/ and Pu/III/ in 1M H2SO4 by Ti/III/, and NaNO2 is chosen to selectively oxidize Pu/III/ and the excess of Ti/III/. The unreacted NaNO2 is destroyed by sulphamic acid and excess Fe/III/ is added following dilution. The equivalent amount of Fe/II/ thus liberated is titrated against standard K2Cr2O7. R.S.D. obtained for the determination of uranium /1–2 mg/ is 0.3% with plutonium being present upto 4.0 mg.

Restricted access

Abstract  

During the simultaneous extraction of plutonium and uranium using anion exchange chromatographic technique from analytical waste in hydrochloric acid medium, 241Am which is invariably present in the plutonium bearing fuel samples remains in the effluent. A two step separation scheme was developed for the recovery and purification of Am from the assorted waste to facilitate the disposal of large volume of aqueous waste and the purified Am solution was utilized for spectroscopic investigation. The separation scheme involved solvent extraction using 0.1 M TODGA + 0.5 M DHOA for separation of americium from Fe, Pb, Ni and Na followed by extraction chromatographic technique using CMPO on inert support as stationary phase for separation of Ca from Am. A systematic study on the extraction behavior of Am from hydrochloric acid medium revealed that out of four extraction systems well known for actinide partitioning namely 0.1 M TODGA + 0.5 M DHOA, 1 M DMDBTDMA, 0.2 M CMPO + 1.2 M TBP and 30% TRPO, only 0.1 M TODGA + 0.5 M DHOA extracts americium from 7.5 M HCl feed acidity. A comparative study involving CMPO solvent extraction and column chromatographic technique revealed that elution of Am from column is satisfactory as compared to inefficient stripping of Am from organic phase in solvent extraction technique using 0.1 M HNO3. The purity of the final solution was checked for 17 elements of interest and was found to be 98% pure, while the overall recovery of this two step separation scheme was found to be 95%.

Restricted access

Abstract  

Energy dispersive X-ray fluorescence (EDXRF) spectrometric methods have been developed for the determination of some common metallic impurities in ThO2 matrix. A series of ThO2 standards containing the analyte impurities in the range 10–100 (or 100–500) μg/g was prepared synthetically. The spectrometer conditions were optimized to obtain calibration plots for the various analytes. The accuracy and precision of the developed methodology for regular assay of ThO2 was evaluated by analyzing three synthetic samples. Further three secondary ThO2 standards were analyzed by EDRXF to check the developed methods. The determined concentrations of Ca, Cr, Fe, Ni and Cu were in good agreement with the certified values of the secondary standards.

Restricted access

Abstract  

Efficacy of chlorinated cobaltdicarbollide in a modified diluent, 20% nitrobenzene in xylene was tested for the extraction and recovery of Cs from simulated high-level waste (HLW) solutions generated from PHWR-fuel reprocessing. Concentration of the reagent, composition of the diluent, numbers of contacts, the nature of stripping agents are some of the parameters optimized for the complete removal of Cs from such waste solutions. The above solvent extraction procedure can be applied to genuine HLW solutions for effective reduction of the dose due to Cs so that HLW can be handled in fume hoods for its characterization.

Restricted access

Abstract  

The high level waste (HLW) generated from the reprocessing of the spent fuel of pressurized heavy water reactor has been characterized for the minor actinides. The radiation dose of the waste solution was reduced by radiochemical separation of cesium from HLW by solvent extraction with chlorinated cobalt dicarbollide dissolved in 20% nitrobenzene in xylene. Minor actinides (Np, Pu, Am, Cm) in the high level waste were assayed by alpha spectrometry following radiochemical separation. The gross alpha activity determined by liquid scintillation agrees well (within 10%) with the cumulative quantities of actinides determined by alpha spectrometry.

Restricted access