Influence of Temperature on the Corrosion of Uranium Dioxide Nuclear Fuel
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0985-NN01-11
Influence of Temperature on the Corrosion of Uranium Dioxide Nuclear Fuel Michael E. Broczkowski, Jamie J. Noel, and David W. Shoesmith Chemistry, The University of Western Ontario, 1151 Richmond St., London, N6A 5B7, Canada ABSTRACT The anodic dissolution of UO2 has been studied at 60ºC and the results compared to previous observations at 22ºC. The rate of oxidation / dissolution was determined electrochemically at constant potentials in the range -500 mV to 500 mV (vs. SCE). The composition of the electrochemically oxidized surface was determined by X-Ray Photoelectron Spectroscopy (XPS). The onset of oxidation (UO2 → UO2+x) occurred at approximately the same potential (-400 mV) at both temperatures. However, the conversion of UV to UVI, and hence to soluble UO22+, was accelerated by temperature. This acceleration of dissolution caused the development of acidity at localized sites on the fuel surface at lower (less oxidizing) potentials (≥ 100 mV) at 60ºC than at 22ºC (≥ 350 mV). INTRODUCTION A considerable effort has been expended on the determination of the mechanism of nuclear fuel (UO2) dissolution under permanent waste disposal conditions [1-11]. These studies show that under the slightly alkaline conditions anticipated, oxidation and dissolution occurs in a sequence of steps, UO2
UO2+x
(UO22+)solution
UO3·yH2O
(1)
with UO2+x being a thin oxidized layer formed prior to dissolution and UO3·yH2O a deposit on the oxidized surface. In our previous studies we have used X-Ray Photoelectron Spectroscopy (XPS) to establish the fractions of UIV, UV and UVI in the UO2 surface after electrochemical and chemical oxidation [12-14]. These studies showed that at low potentials the UO2 surface is oxidized to UO2+x by O2⎯ injection into the UO2 fluorite lattice leading to the conversion of UIV to UV. At more oxidizing applied potentials, dissolution as UO22+ leads to the deposition of a UVI deposit (most likely UO3·yH2O in dilute saline solution [1]). In silicate-containing solution a hydrated UVI silicate was formed [14]. At extremely oxidizing potentials (> 0.4 V), the rapid formation and hydrolysis of UO22+ led to local acidification in pores in the deposited UO3·yH2O layer and flaws in the fuel surface, and a reactivation of the fuel surface [15]. This reactivation was attributed to the re-dissolution of the UO3·yH2O deposit and a partial dissolution of the underlying UO2+x surface layer. In this paper, we present the results of electrochemical and surface analytical (XPS) studies on the anodic dissolution of UO2 at a temperature of 60ºC. This temperature was chosen to mimic those anticipated on fuel surfaces inside a failed nuclear waste container. Calculations of the thermal evolution within a waste container show that the fuel temperature will be in the range 80ºC to 40ºC for the majority of the assumed failure scenario [16]. The results are compared to previous data at room temperature (~ 22ºC) [1]. Our primary goal is to determine how temperature influences the extent of oxidation of the fuel surfa
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