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Abstract  

The present method describes the determination of hydrazine by making use of potentiometric titration technique. The underlying principle is back titration of uncreacted excess cerium remaining after the complete oxidation of hydrazine. Standardized ferrous ammonium sulfate was used for titration. This method was applied to real samples generated from a nuclear reprocessing plant wherein control of hydrazine is of paramount importance. The interference of U(IV), Cr(III), U(VI), nitrite, and chloride was studied and of all these ions the way to eliminate the interference of U(IV) was only attempted. The relative standard deviations (RSD) for synthetic as well as real samples were determined. The method gives RSD of less than 1% in the range of 1 mg to 20 mg of hydrazine. The error in the range 3 mg to 17 mg was found to be less than 1%.

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Abstract  

A very sensitive extraction spectrophotometric method for the analysis of uranium based on the extraction of a uranium—benzoate—crystal violet complex by a mixture of xylene and benzene is described. The absorbance maximum is at 606 nm and molar absorptivity is 4.28·104 l·mol−1·cm−1. The interference due to a number of anions and cations studied without any pre-extraction was found to be within permissible limits. The method has been used for determining uranium in a synthetic solution, i.e., uranium in the presence of various other ions. The interference due to some cations was eliminated by the use of a masking agent (boric acid).

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Abstract

The square-pyramidal copper(II) complexes with ciprofloxacin in the presence of bipyridine derivatives have been synthesized and characterized using elemental analysis, magnetic moment measurement, thermal analysis (TG), IR, mass and reflectance spectra. The thermal denaturation study has been used for evaluating calf thymus DNA interaction activity. Spectral and hydrodynamic measurements have been used for validating the DNA interaction study. The thermodynamic profile was established for proper understanding of DNA binding Gibbs free energy.

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Summary

The present study reports the dissolution method for a novel fixed dose combination (FDC) containing etodolac (ET) and propranolol hydrochloride (PH) developed utilizing USP Apparatus 1 (basket) at 100 rpm with 1000 mL of phosphate buffer (pH 6.8; 0.05 M) medium at 37°C. An isocratic reversed-phase liquid chromatographic (RPLC) method was also developed for the simultaneous determination of ET and PH on an octadecylsilica column using phosphate buffer (pH 5.5) and acetonitrile (60:40, υ/υ) as the mobile phase with ultraviolet (UV) detection at 292 nm. Validation data were obtained, which demonstrated that the dissolution methodology is accurate, precise, linear, and rugged for the combination tablets.

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Abstract  

The spent fuel from Fast Breeder Test Reactor of various burnups from 25 to 155 GWd/te is being reprocessed in CORAL (COmpact Reprocessing of Advanced fuels in Lead shielded cell) using a modified PUREX (Plutonium Uranium Recovery by EXtraction) process. Total plutonium (Pu238, 239, 240, 241 & 242) concentration in the sample is analysed by HTTA (Thenoyl Trifluoro Acetone) extraction method wherever interference from other alpha emitting nuclides (Raffinate) and bulk natural uranium (uranium products) are present "as reported by Milyukov et al. (Analytical chemistry of plutonium, <cite>1967</cite>) and Natarajan and Subba Rao (BARC, pp. 38–43, <cite>2007</cite>)". This method requires the addition of corrosive reagents such as NH2OH.HCl which is a problem in waste disposal for reduction. A salt-free reagent such as Hydroxyurea is studied as a reducing agent which has the ability to reduce both Pu(VI) and Pu(IV) to Pu(III) "as reported by Zhaowu (260(3):601–606, <cite>2004</cite>) and Zhaowu (262(3):707–711, <cite>2004</cite>)". Pu(III) thus formed can be easily oxidised to Pu(IV) by NaNO2 for the extraction of Pu by HTTA.

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Abstract  

A radiotracer study was carried out in a trickle bed reactor (TBR) independently filled with two different types of packing i.e., hydrophobic and hydrophilic. The study was aimed at to estimate liquid holdup and investigate the dispersion characteristics of liquid phase with both types of packing at different operating conditions. Water and H2 gas were used as aqueous and gas phase, respectively. The liquid and gas flow rates used ranged from 0.83 × 10−7–16.67 × 10−7 m3/s and 0–3.33 × 10−4 m3 (std)/s, respectively. Residence time distribution (RTD) of liquid phase was measured using 82Br as radiotracer and about 10 MBq activity was used in each run. Mean residence time (MRT) and holdup of liquid phase were estimated from the measured RTD data. An axial dispersion with exchange model was used to simulate the measured RTD curves and model parameters (Peclet number and MRT) were obtained. At higher liquid flow rates, the TBR behaves as a plug flow reactor, whereas at lower liquid flow rates, the flow was found to be highly dispersed. The results of investigation indicated that the dispersion of liquid phase is higher in case of hydrophobic packing, whereas holdup is higher in case of hydrophilic packing.

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Abstract  

Ammonium uranyl carbonate (AUC) precipitation is developed for the conversion of uranyl nitrate to oxide in the uranium reconversion step of reprocessing of irradiated fuel by the addition of ammonium carbonate salt. Different precipitation conditions of AUC are studied. The solubility of AUC as a function of uranium concentration in the feed at different temperatures using ammonium carbonate salt as precipitant is studied. This study indicates that 95-99.8% of uranium is recovered as AUC by precipitating 5-125 g/l of uranium with loss of uranium (250-10 ppm) in the filtrate by adding ammonium carbonate salt. It is also observed that the solubility of AUC increased as the concentration of uranium decreased. Thermal decomposition is carried out by thermogravimetry/differential thermal analysis (TG/DTA) and evolved gas analysis-mass spectrometry (EGA-MS) to find out AUC decomposition and gases evolved during decomposition. Studies are also carried out to characterize AUC by using X-ray diffraction (XRD). The data show that AUC obtained by the above conditions is very much consistent with published information.

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Abstract  

A method is developed for the selective leaching of 233U from a thorium oxalate cake. The leaching capacity of ammonium carbonate and nitric acid have been investigated, showing that (NH4)2CO3 leads to higher recovery. The maximum leaching efficiency is obtained using 0.5% ammonium carbonate, with a minimal thorium pick-up. A uranium recovery of 94% is obtained after three consecutive contact experiments in carbonate media, with minimal thorium uptake in the leachate. This process was applied to an actual plant stream, allowing the reduction of the 233U -activity from 5.64 to 0.3 Ci/g of thorium oxalate cake.

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Abstract  

A two step precipitation using ammonium carbonate and oxalic acid as the precipitants for thorium and iron is developed for the purification of 233U. Ammonium carbonate is added to the feed to increase the pH of the solution. The effect of pH on the solubility of U, Th and Fe in an excess of ammonium carbonate is studied. This indicates that the solubility of Th and Fe is minimum at pH 7 and the recovery of uranium is maximum. The effect of the concentration of thorium and iron on the recovery of uranium at pH 7 is studied. This indicates that the ammonium carbonate precipitation tolerates 2 g/l of thorium and 10 g/l of iron keeping losses of uranium to a minimum. If the feed solution contains more than a tolerable concentration of thorium the precipitation is followed in two steps: (1) Bulk of the thorium is removed by oxalate precipitation, (2) the remaining thorium and iron in the supernatant are removed by ammonium carbonate precipitation. A flow sheet is proposed for the purification of 233U from thorium and iron present in a strip product concentrate obtained during the reprocessing of irradiated thorium rods.

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Journal of Radioanalytical and Nuclear Chemistry
Authors:
A. Palamalai
,
S. Mohan
,
M. Sampath
,
R. Srinivasan
,
P. Govindan
,
A. Chinnusamy
,
V. Raman
, and
G. Balasubramanian

Abstract  

Some batches of233U oxide product obtained from the reprocessing treatment of irradiated thorium rods, called J-rods in our plant, have been found to contain thorium as much as 85% and iron above 5% as impurities. This product has to be purified before sending for fabrication of the fuel. The present purification method consists of the following three steps: (1) preferential dissolution of U3O8 as compared to thoria, (2) a novel solvent extraction process, and (3) preferential precipitation of Th as oxalate leaving behind the entire U in the filtrate. Development and application of the present purification method to the above233U oxide proxduct are presented in this paper.

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