R. V. Subba Rao

Determination of Uranium and Plutonium in High Active Solutions by Extractive Spectrophotometry

Plutonium and uranium was extracted from nitric acid into trioctyl phosphine oxide in xylene. The TOPO layer was analysed by spectrophotometry. Thoron was used as the chromogenic agent for plutonium. Pyridyl azoresorcinol was used as chromogenic agent for uranium. The molar absorption coefficient for uranium and plutonium was found to be 19000 and 19264 liter/mole-cm, respectively. The correlation coefficient for plutonium and uranium was found to be 0.9994. The relative standard deviation for the determination of plutonium and uranium was found to be 0.96% and 1.4%, respectively.

Optimization of Flowsheet for Scrubbing of Ruthenium during the Reprocessing of Fast Reactor Spent Fuels

The solvent extraction method using tri-n-butyl phosphate diluted with heavy normal paraffin is being used as an extractant for the recovery of plutonium and uranium present in spent fuel discharged from a nuclear reactor. The purification of uranium and plutonium from the fission product ruthenium is considered very important in the fast reactor fuel reprocessing process due to its complex chemistry. Systematic studies have been carried out in batch as well as countercurrent solvent extraction for optimization of parameters to get maximum decontamination factor for ruthenium. In dual scrubbing condition using 3 M and 6 M nitric acid with 60–70% uranium loading in organic phase, a decontamination factor of >103 for Ru is obtained in mixer-settler run.

Effect of Aliquat 336 concentration on transportation of hydrochloric acid across supported liquid membrane

The transport of hydrochloric acid across a supported liquid membrane using Aliquat 336 in xylene as a carrier was studied. The effect of carrier concentration (0.1–0.6M) on the transportation of hydrochloric acid with and without phase modifier was investigated. The study indicated that the flux of transportation decreased with increasing carrier concentration in the absence of phase modifier. In the presence of phase modifier, however, the flux increased up to 0.2M carrier concentration and started decreasing afterwards. The transportation behavior of hydrochloric acid with and without phase modifier has been attributed to the tendency of aggregation of the carrier.

Purification of uranium product from plutonium contamination using acetohydroxamic acid (AHA) based process

Acetohydroxamic acid (AHA) based uranium product purification process to remove plutonium was optimized. For this process, equilibrium data was generated to optimize AHA concentration and acidity of stripping agent/scrubbing agent. Two options namely (i) Pu complexation in aqueous phase followed by extraction and scrubbing ii) extraction followed by scrubbing with AHA were studied. Results of these studies indicate that U product obtained in AHA purification is near to the table top specification and also quantitative Pu recovery from the AHA strip product is possible by oxalate precipitation.

Partitioning of plutonium and uranium in aqueous medium using hydroxyurea as reducing agent

A new process for the partitioning of plutonium and uranium during the reprocessing of spent fuel discharged from fast reactor was optimised using hydroxyurea (HU) as a reductant. Stoichiometric ratio of HU required for the reduction of Pu(IV) was studied. The effect of concentration of uranium, plutonium and acidity on the distribution ratio (Kd) of Pu in the presence of HU was studied. The effect of HU in further purification of Pu such as solvent extraction and precipitation of plutonium as oxalate was also studied. The results of the study indicate that Pu and U can be separated from each other using HU as reductant.

Purification of 233U from Thorium and Iron in the Reprocessing of Irradiated Thorium Oxide Rods

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.

Ammonium uranyl carbonate (AUC) based process of simultaneous partitioning and reconversion for uranium and plutonium in fast breeder reactors (FBRs) fuel reprocessing

Ammonium uranyl carbonate (AUC) based process of simultaneous partitioning and reconversion for uranium and plutonium is developed for the recovery of uranium and plutonium present in spent fuel of fast breeder reactors (FBRs). Effect of pH on the solubility of carbonates of uranium and plutonium in ammonium carbonate medium is studied. Effect of mole ratios of uranium and plutonium as a function of uranium and plutonium concentration at pH 8.0–8.5 for effective separation of uranium and plutonium to each other is studied. Feasibility of reconversion of plutonium in carbonate medium is also studied. The studies indicate that uranium is selectively precipitated as AUC at pH 8.0–8.5 by adding ammonium carbonate solution leaving plutonium in the filtrate. Plutonium in the filtrate after acidified with concentrated nitric acid could also be precipitated as carbonate at pH 6.5–7.0 by adding ammonium carbonate solution. A flow sheet is proposed and evaluated for partitioning and reconversion of uranium and plutonium simultaneously in the FBR fuel reprocessing.

Development of a reconversion method for uranyl nitrate to oxide in the reconversion step of reprocessing of irradiated fuel

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.

Improved Preparation of Acetohydroxamic Acid

Acetohydroxamic acid (AHA) is a good chelating agent for heavy metals1 such as copper, iron, cobalt, nickel, chromium, manganese, uranium. It has been suggested as a partitioning agent for the separation of uranium and plutonium in spent nuclear fuel reprocessing.2–4 Thus a simple and economical preparation of AHA is of importance. The basic procedure for the preparation of hydroxamic acids involves the reaction of esters with hydroxylamine hydrochloride in the presence of an alkali, the free acetohydroxamic acid being liberated by the addition of a calculated quantity of a mineral acid to the cold reaction mixture (Equation (1)).5

Development of a method for recovery of 233U from thorium oxalate cake in reconversion step of reprocessing of irradiated thorium rods

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 α-activity from 5.64 to 0.3 μCi/g of thorium oxalate cake.