Authors:P. Govindan, A. Palamalai, K. Vijayan, M. Raja, S. Parthasarathy, S. Mohan, and R. Rao
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.
Authors:A. Palamalai, S. Mohan, M. Sampath, R. Srinivasan, P. Govindan, A. Chinnusamy, V. Raman, and G. Balasubramanian
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.
Authors:P. Govindan, A. Palamalai, K. Vijayan, S. Subbuthai, S. Murugesan, S. Mohan, and R. Subba Rao
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.
Authors:P. Govindan, A. Palamalai, K. Vijayan, K. Dhamodharan, S. Subbuthai, S. Mohan, and R. Subba Rao
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.
Authors:Rajesh Rajan, N. P. Padmaja, V. Ramakrishna Pillai, Rachel Daniel, and Govindan Vijayaraghavan
Ventricular septal rupture (VSR) is a rather rare, but at the same time very dreadful complication of acute myocardial infarction in the percutaneous coronary intervention (PCI) era and only limited evidence exist on the optimal treatment of this critical medical condition. VSR is less common following successful early thrombolysis and PCI occurring in myocardium supplied by infarct-related artery (IRA). We report two well-documented cases of successful VSR treatment which will provide valuable information for clinical practice especially due to the long-tem follow-up. Both cases underwent delayed elective surgical closure of VSR. This report clearly describes the incidence, potential risks and timing of occurrence, clinical features, and outcomes of ventricular septal rupture complicating acute myocardial infarction (AMI) after PCI. Hence the topic of this report is of great importance. Although the prognosis of patients who develop VSR is generally grave without immediate surgical repair, both our patients remained hemodynamically stable at discharge and during follow-up of more than 4 years.
Authors:P. Govindan, S. Sukumar, K. Vijayan, G. Santhosh Kumar, S. Ganesh, Pradeep Sharma, K. Dhamodharan, R. Subba Rao, M. Venkataraman, and R. Natarajan
A novel method has been developed for recovery of plutonium and uranium from carbonate wash solutions generated during solvent
wash process involved in the reprocessing of high burn up FBTR fuel. The proposed method involves a selective coprecipitation
of Pu and U by adding ammonium hydroxide to the pre acidified carbonate wash solution. Substantial removal of DBP by successive
steps of coprecipitation, completely eliminates the possibility of undesired solid formation which is mainly due to the presence
of high content of DBP. By adopting this method, an excellent decontamination factor for DBP has been achieved without any
crud/solid formation. Phosphate content in the final oxide product meets the product specifications. Flowsheet condition necessary
for the recovery process for plutonium from the aqueous carbonate solution is formulated and adopted in the CORAL facility.