Purification of uranium from uranium ore concentrates dissolved in ammonium carbonate and hydrogen peroxide to be used as a feed stock material for a new conversion plant

Abstract

Enriched Uranium (EU), the primary material used to fabricate fuel for nuclear power plants, is manufactured using uranium oxide as feed material. Currently South African gold mines produce tonnes of uranium as a byproduct during gold extraction and this uranium byproduct ore converted at the Nuclear Fuel Corporation of South Africa (Nufcor) into uranium oxide (UOC) for example triuranium octaoxide (U3O8) . However other possible sources (gold mines from international suppliers) of uranium oxide for example U3O8 or UO2 could contain impurities that will cause problems in a conversion plant as experienced previously in a conversion plant at Necsa (1976 - 1979). Impurities in the ammonium diuranate (ADU) as feed material resulted in the formation of sintered UO2F2 particles during the conversion process (Ponelis et al., 1987). Purification of UOC is therefore, required to ensure a continuous pure UOC feed for conversion plants. This study deals with the development and optimization of an ammomum carbonate ((NH4)2CO3) based dissolution process using hydrogen peroxide (H2O2) as an oxidant for UOC and the purification of uranium from the generated solutions. The dissolution experiments were performed in a pressurized autoclave as well as bench experiments. Experimental parameters that were investigated for optimum dissolution efficiency; include addition of H2O2, temperature, solid/liquid ratio, H2O2 concentration, (NH4)2CO3 concentration and stirring rate. The reaction time required for optimum dissolution was also determined. The results obtained indicated a complete dissolution of UOC at 60 C, after three hours in a 1 M ammonium carbonate solution with 1 M hydrogen peroxide. The rate of the reaction and the yield of uranium were found to increase as a function of both the concentration of hydrogen peroxide in the range of 0.5 to 2.5 Mand the temperature between 30 and 60°C. Three purification processes were investigated to purify the generated uranyl solution, namely ion exchange resins, solvent extraction and precipitation. For the ion exchange technique, various resins were tested for the efficiency of absorption of the impurities such as potassium, sodium, tungsten, molybdenum, calcium and aluminium from solutions containing high concentrations of uranium. Results to be presented indicated that Amberlite RFH 252 and Purolite S957 could be considered for purification purposes. For the solvent extraction technology, the possibilities of purifying uranium from impurities in an alkaline medium using Aliquat 336 or TBP as extractants in the presence of xylene, dodecane and kerosene as diluents were studied. Results indicated that, although uranium extraction was possible from an alkaline medium using Aliquat 336 in xylene, impurities were also extracted. Precipitation technology (steam stripping) was also investigated for selective precipitation of uranium. Results indicated possible and feasible recovery of uranium by steam stripping between 70 and 90 C. Purification of uranium feed stock for a conversion plant from different sources (which might contain impurities) as a three steps approach was recommended. The first step is the dissolution of uranium feed stock by alkaline solution followed by purification by 10n exchange using Amberlite RFH 252 or Purolite S 957. In the third step, the eluent must be precipitated and the precipitates need isolation for feed material. The recommended solution consists of a combination of ion exchange and liquid separation technologies.