Search Results

You are looking at 1 - 4 of 4 items for

  • Author or Editor: A. Palamalai x
  • Refine by Access: All Content x
Clear All Modify Search
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.

Restricted access

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.

Restricted access

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.

Restricted access

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.

Restricted access