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- Author or Editor: S. Tripathi x
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Abstract
The techniques of gamma-radiolysis, UV photolysis and hydrogen-induced reduction of aqueous palladium perchlorate to ultrafine particles of Pd, in the presence of alumina sol, have been studied. As compared to H2-induced reduction, both UV photolytic and gamma-radiolytic reduction methods lead to a very stable, brown colored Pd colloid with relatively less absorption in the higher wavelength region and possessing much smaller volume average particle size (62 and 61 nm, respectively). Higher concentration of alumina sol and increase in pH from 1.8 to 7.2, favour the formation of smaller sized particles as determined by the dynamic light scattering technique.
Abstract
Present investigation describes the development of a spectrophotometric method for trace level determination of U(VI) encountered during the process of nuclear fuel fabrication and reprocessing industries. A chromogenic reagent, 2-(5-bromo-2-pyridylazo-5-diethylaminophenol) (Br–PADAP) is used to complex with U(VI) under optimized solution conditions. The absoption maxima of the uranyl Br–PADAP complex at 578 nm is computed to be 73540 ± 1438 for U–Br–PADAP solution containing 20% ethanol (in aqueous sample media) and 58216 ± 1208 for U–Br–PADAP solution containing 80% ethanol (for organic sample media). Employing suitable sample treatment methods, the scope of analytical method has been widened to permit accurate determination of U(VI) in the samples with variation in the relative compositions of Th(IV), Pu(IV) and Fe(III). The method is applicable to samples matrices with, acidic, alkaline highly salted media. Effect of commonly associated ionic species on the optical density of U–Br–PADAP is determined. Depending on the extent of the interfering impurities present, the method permits estimation of U(VI) either direct or after its selective extraction into tri-octyl phosphine oxide dissolved in cyclohexane. The method is precise with <5% standard error and can be used for the estimation of uranium in organic as well as in aqueous samples. The method has been validated for quantitative determination of uranium extracted in the organic phase comprising of heavy metal extractants like TBP, HDEHP, PC-88A and Aliquot 336.
Abstract
Comprehensive quality assurance/quality control procedure is very much necessary to obtain accurate and precise analytical measurement result. This paper discusses the quality control aspects of the High-Purity Germanium (HPGe) based gamma spectrometry system, which has been used for the measurement of low-level radioactivity in environmental samples. The gamma spectrometry system consisting of coaxial n-type HPGe detector having 50% relative efficiency with respect to 7.62 cm x 7.62 cm NaI (Tl), Nuclear Instrumentation Module (NIM) based pulse processing electronic accessories and 8 k MCA. To reduce the background contribution, 7.5 cm thick lead has been placed surrounding the detector. The minimum detectable activities (MDA) with 95% confidence level (for 300 g soil sample and 100,000 s counting time) for important radionuclides such as 238U, 226Ra, 232Th, 40K, and 137Cs are 10.4, 4.3, 4.1, 16.9 and 0.1 Bq kg−1, respectively. The Quality control (X bar R) charts were plotted using 137Cs and 40K background counts observed periodically, which showed that the fluctuation is well within the confidence limit and confirms the stability of the system. The laboratory has been participating in the proficiency tests (PTs) of the International Atomic Energy Agency (IAEA). In recently concluded PTs, the samples include soil, spiked standard solution, spinach, phosphogypsum and spiked air filter were analysed for the natural, fission and activation products radionuclides. The performance evaluation of the IAEA PTs showed that the laboratory results were in good agreement with the target value, which confirms the reliability and traceability of the gamma spectrometric measurement result of the laboratory.
Abstract
This paper presents a simple, rapid and sensitive radiometric method for the determination of uranium in Thorex Process stream containing large amount of thorium. This method involves the extraction of uranium into 0.05M tri-n-octyl phosphine oxide (TOPO) in xylene at 2M HNO3. The extraction of thorium is prevented by masking them with suitable quantity of fluoride ions. The optimum experimental parameters for extraction of 233U were evaluated and using the most suitable experimental conditions the extracted uranium is radiometrically determined by α-counting in proportional counter with a prior knowledge of specific activity of uranium. Simultaneously in the same sample uranium was determined by spectrophotometric method using 2-(5bromo-2 pyridylazo)-5-diethylaminophenol (Bromo-PADAP) as chromogenic reagents. Simulated as well as actual samples of dissolver, conditioner and raffinate tank of Thorex stream have been analyzed by both these methods. The method was tested for as low as 0.15 μg of uranium and the results of these analyses were found to be satisfactory within the experimental limits.
Abstract
This paper discusses the measurement of naturally occurring radioactivity materials (NORM) in beach sand minerals using high resolution gamma spectrometry. In India, the beach sand minerals of economic interest from coastal Kerala, Tamil Nadu and Orissa are enriched with NORM due to the occurrence of monazite deposits and heavy minerals such as zircon, ilmenite, magnetite, garnet, rutile etc. Since many of these ores are rich in 232Th and other radio elements, certification of radioactivity levels has become mandatory in recent years. The average activity concentrations of 226Ra in zircon, rutile and garnet were 3,531, 1,134 and 17 Bq kg−1, respectively. The average activity concentration of 232Th observed in zircon, rutile and garnet were 618, 454 and 64 Bq kg−1, respectively. Concentration of 226Ra, 232Th, and 40K in ilmenite ore ranged from 17.6–444 Bq kg−1, 80.4–1971 Bq kg−1 and ≤5.5–25.0 Bq kg−1, respectively.
The dithiocarbamato complexes of titanyl(IV), zirconyl(IV) and hafnyl(IV), abbreviated as MO(S2CNRR)2·nH2O(M=Ti, Zr or Hf,R=H,R′=C5H9;R=H,R′=C7H11,n=1 for Ti andn=2 for Zr and Hf), were prepared in aqueous medium and characterized by elemental analyses, magnetic susceptibility measurements and IR spectral studies. The thermal behaviour of these compounds under non-isothermal conditions was investigated by thermogravimetric, derivative thermogravimetric and differential scanning calorimctric techniques in nitrogen and oxygen atmospheres. The intermediates obtained at the end of various thermal decomposition steps were identified on the basis of analyses and IR spectral studies. Kinetic parameters, such as apparent activation energy and order of reaction, were determined by the graphical method of Coats and Redfern. The heats of reaction for the different decomposition steps were calculated from the DSC curves.
Abstract
2-Ethylhexyl-2-ethylhexyl phosphonic acid (PC-88A) and Tributylphosphate (TBP) extractants have been attached to polypropylene (PP) in granular, film and non-woven fabric forms, by a simultaneous γ-ray irradiation method. The extraction of plutonium from the acidic radioactive liquid waste by modified polymers was studied by varying the γ-ray dose. The uptake of plutonium was also studied by polyethylene (PE) in film form. This modified polymer also showed extraction capability for plutonium from nitric acid medium. The uptake of plutonium depends upon the γ-ray dose as well as the nature and source of the polymer. Liquid–solid extraction studies showed that the equivalent amount of uptake of plutonium on TBP–polyethylene film requires twice the γ-ray dose as compare to TBP–polypropylene film. It was observed that at given γ-ray dose polypropylene fabric is not sturdy, compare to the granules and films, and material leach out in aqueous phase. The presence of other solvents like di-methyl formamide (DMF) and cyclohexane during γ-ray irradiation were able to enhance the extraction capabilities. The optimum conditions established during this study was successfully applied for the separation of plutonium, uranium and thorium from the fission products in acidic waste solution.
Abstract
An emulsion liquid membrane (ELM) containing di-2-ethylhexylphosphoric acid (D2EHPA) as the carrier extractant and SPAN 80 as the surfactant was used to pre-concentrate Am3+ from dilute acid solutions. Effects of various factors such as: external phase pH, internal phase conditions, equilibration time, D2EHPA concentration, SPAN 80 concentration, etc. on Am3+ mass transfer were investigated. Emulsion was broken by the addition of solvents such as acetone and the actual mass transfer obtained after breaking the emulsion agreed well with that obtained by the difference method.
Abstract
This paper describes the studies on the extraction of molybdenum (VI) from aqueous nitric acid medium by (2-ethylhexyl) phosphonic acid, mono (2-ethylhexyl) ester (PC-88A). The extraction affecting parameters such as concentration of HNO3 in aqueous feed, effect of concentration of extractants, effect of diluents, and molybdenum concentration in the aqueous phase are investigated to optimize the extraction conditions for the quantitative separation of molybdenum from nitric acid medium. With increase of HNO3 concentration in aqueous phase, percentage extraction was found to be decreased in all the cases. Percentage extraction of molybdenum increases with increase in PC-88A concentration till the 0.15 M of PC88A, and after that it becomes constant. Kerosene and n-dodecane was found to be most suitable diluents. Among the various strippants used 0.2 M (w/v) solution of Na2CO3 and 0.2 M (w/v) solution (NH4)2CO3 are found to be the equally suitable for stripping of molybdenum from the loaded organic phase. The stripping of molybdenum from loaded organic layer by various reagents followed the order: (NH4)2CO3 >Na2CO3 >0.1 M sodium salt of EDTA >2 M NaOH >8 M HNO3. The optimized process conditions are employed to extract molybdenum (VI) from actual Davies–Gray waste as well as from diluted high level waste generated in the purex stream. More than 94% Mo(VI) was extracted from radioanalytical as well as from high level waste of purex process and quantitative recovery was achieved in both the cases when 0.2 M sodium carbonate was used as stripping agent.
Abstract
This paper deals with the studies on decontaminations of spent ion exchange resin used for purification of plutonium in PUREX process stream. Studies were carried out to optimize the chemical procedure for removal of plutonium and fission products activities form spent Ion Exchange resin. Different metal complexing reagents were tested for leaching out of radionuclides entrapped in irradiated spent ion exchange resin. The experimental results indicate that 0.01 M NaF solution was found the most suitable for removal of plutonium. The mixture of Na2CO3 and sodium salt of EDTA solution was found to be better for decontamination of spent ion exchange resin from beta and gamma activities. Optimized mixture of 0.5 M Na2CO3 and 0.1 M sodium salt of EDTA solution was found to be the most effective for fission product activities removal. After successive multiple contacts using these suitable reagents, the Pu and fission product activities in spent ion exchange resin were brought down to a minimum possible level, making it quite suitable for its long term storage.
