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  • Author or Editor: P. Pathak x
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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.

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Abstract  

Extraction of uranium from tissue paper, synthetic soil, and from its oxides (UO2, UO3 and U3O8) was carried out using supercritical carbon dioxide modified with methanol solutions of extractants such as tri-n-butyl phosphate (TBP) or N,N-dihexyl octanamide (DHOA). The effects of temperature, pressure, extractant/nitric acid (nitrate) concentration, and of hydrogen peroxide on uranium extraction were investigated. The dissolution and extraction of uranium in supercritical CO2 modified with TBP, from oxide samples followed the order: UO3 ≫ UO2 > U3O8. Addition of hydrogen peroxide in the modifier solution enhanced the dissolution/extraction of uranium in dynamic mode. DHOA appeared better than TBP for recovery of uranium from different oxide samples. Similar enhancement in uranium extraction was observed in static mode experiments in the presence of hydrogen peroxide. Uranium estimation in the extracted fraction was carried out by spectrophotometry employing 2-(5-bromo-2-pyridylazo)-5-diethylaminophenol (Br-PADAP) as the chromophore.

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Abstract  

Polyacrylhydroxamic acid resin synthesized by functionalization of polyacrylamide with hydroxylamine has been investigated for the sorption of plutonium(IV) from carbonate medium, aiming at its application for the removal of plutonium from alkali wash effluent generated during purification of TBP in PUREX process. Batch experiments have been carried out to determine distribution coefficient of plutonium(IV) between this exchanger and various compositions of carbonate medium. Effect of the concentration of sodium carbonate, sodium bicarbonate and pH of the solution on the distribution coefficient have been studied to optimize the conditions for the uptake of Pu(IV) by this exchanger. Column experiments were carried out to determine the practical capacity of the exchanger for plutonium. Elution studies were also carried out to recover the loaded plutonium from the ion exchange column The exchanger displayed good exchange capacity for Pu(IV) from feed solution simulating the conditions of carbonate wash effluent generated in PUREX process. The exchanger also exhibited fast elution of Pu, suggesting the feasibility of using it for the recovery of Pu from carbonate based wash effluent.

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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.

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Abstract  

A method for the separation of Fe(III) from Co(II) in nitric acid medium has been developed using 30% Cyanex-923 in n-dodecane. Nitric acid extraction studies (0.5 to 6.0M) as a function of Cyanex-923 concentration (10-30%) suggest the probable extraction of more than one species of Cyanex-923. HNO3 complex. Third phase formation was observed at 8.0M HNO3. Extraction of Fe(III) increases monotonously with acidity (1.0-7.0M) whereas that of Co(II) is negligible at all acidities. Fe(III) was quantitatively loaded in the organic phase in two contacts at 7.0M HNO3

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Abstract  

Sorption behavior of 241Am (~10−9 M) on naturally occurring mineral pyrite (particle size: ≤70 μm) has been studied under varying conditions of pH (2–11), and ionic strength (0.01–1.0 M (NaClO4)). The effects of humic acid (2 mg/L), other complexing anions (1 × 10−4 M CO3 2−, SO4 2−, C2O4 2− and PO4 3−), di- and trivalent metal ions (1 × 10−3 M Mg2+, Ca2+ and Nd3+) on sorption behavior of Am3+ at a fixed ionic strength (I = 0.10 M (NaClO4)) have been studied. The sorption of 241Am on pyrite increased with pH from 2.8 (84%) to 8.1 (97%). The sorption of 241Am decreased with ionic strength at low pH values (2 ≤ pH ≤ 4), but was insensitive in the pH range of 4–10, suggesting the formation of outer-sphere complexes on pyrite surface at lower pH, and inner-sphere complexes at higher pH values. The sorption of 241Am increased in the presence of (i) humic acid (5 < pH < 7.5), and (ii) C2O4 2− (2 < pH < 3). By contrast, other complexing anions such as (carbonate, phosphate, and sulphate) showed negligible influence on 241Am sorption. The presence of Mg2+, Ca2+ ions showed marginal effect on the sorption profile of 241Am; while the presence of Nd3+ ion suppressed its sorption significantly under the conditions of present study. The sorption of 241Am on pyrite decreased with increased temperature indicating an exothermic process.

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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.

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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.

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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.

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Abstract  

The extraction of uranium(VI) from nitric acid medium is investigated using 2-ethylhexyl phosphonic acid-mono-2-ethylhexyl ester (PC88A in dimeric form, H2A2) as extractant either alone or in combination with neutral extractants such as tri-n-butyl phosphate (TBP), trioctyl phosphine oxide (TOPO), and dioctyl sulfoxide (DOSO). The effects of different experimental parameters such as aqueous phase acidity (up to 10 M HNO3), nature of diluent [xylene, carbon tetrachloride (CCl4), n-dodecane and methyl iso-butyl ketone (MIBK)] and of temperature (303–333 K) on the extraction behavior of uranium were investigated. Synergistic extraction of uranium was observed between 0.5 and 6 M HNO3. Use of MIBK as diluent was also studied. Temperature variation studies using PC88A as extractant showed exothermic nature of extraction process. Studies were carried out to optimize the conditions for the recovery of uranium from the raffinate generated during the purification of uranium from nitric acid medium. Inductively Couple Plasma Atomic Emission Spectroscopy (ICP-AES) and Energy Dispersive X-Ray Fluorescence (EDXRF) techniques were employed for analysis of uranium in equilibrated samples.

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