The Surface Precipitation During UO 2 Leaching Process
- PDF / 533,041 Bytes
- 6 Pages / 595 x 842 pts (A4) Page_size
- 99 Downloads / 242 Views
II9.1.1
THE SURFACE PRECIPITATION DURING UO2 LEACHING PROCESS Daqing Cui1 , Jérôme Devoy1 and Kastriot Spahiu2 1 2
Studsvik Nuclear AB, SE-611 82 Nyköping, Sweden SKB, P.O. Box 5864, SE-10240 Stockholm, Sweden
To understand the influence of cations and silica in groundwater on spent fuel corrosion, the leaching behavior of newly reduced UO2.00 in different solutions was investigated. Uranium concentrations in the solutions were measured and the leached UO2 surface and precipitates on it were analyzed by XPS, laser Raman spectroscopy, SEM-EDS and TEM-EDS. It was found that, in air saturated 2 mM HCO3- solutions, the leaching rates of UO2(s) depend on the composition of the leaching solutions. Very similar leaching rates, 0.5 µg • cm-2 • day-1, were obtained in 0.1 M NaCl and 0.1 M KCl solutions, while that in 0.45mM Ca2+ -0.18 mM Mg2+ -0.2 mM SiO2 containing synthetic groundwater was 10 times smaller. Calcite and quartz like precipitates were detected on the corroded UO2(s) surface. INTRODUCTION The geological repositories of spent nuclear fuel are being considered either in an air saturated zone or in deep hard rock depending on the geological conditions of different countries. Independently of the options chosen, the kinetics of dissolution of spent fuel and UO2(s) under oxic groundwater conditions has been studied extensively [1-5], since the radiolysis creates oxidative conditions near the fuel surface even when the groundwaters are reducing. Spent nuclear fuel is largely UO2(s), with only a small fraction of transuranium elements and fission products. Most radionuclides are present as solid solution within the fuel matrix and their release rates depend mainly on the rate of dissolution of the UO2(s) fuel matrix. Hence studies of uranium dioxide dissolution contribute to an increased understanding of the spent fuel matrix alteration/dissolution process. It is well known that HCO3- is a strong complexing agent of U(VI) that plays an important role in the UO2(s) leaching kinetics and U(VI) solubility. The leaching rate of spent fuel depends very much on the chemical composition of the leaching solution. In previous spent fuel leaching experiments in this laboratory [2] with oxic synthetic (Allard) groundwater containing 2 mM HCO3-, the total uranium leached into solution increased to 4 • 10-6 M over the first week, then remained practically at constant levels for ~500 days. At longer cumulative contact times, there is a slow increase in concentration, while after about a thousand days, the U concentration reached 2 •10-5 M, which is much lower than the predicted solubility of U(VI) minerals in 2 mM HCO3- solution. A similar observation was made by Wilson [1]. Forsyth and Werme explained this observation by the formation of U3O7/ U3O8 redox buffer in the air present leaching system and found the calculated solubility of U3O7/ U3O8 to match the U(VI) concentration obtained in the long time leaching experiment [2]. In a more recent investigation [6], equilibrium cation exchange reactions involving NH4+, Na+, K+, H+
Data Loading...
