D.W. Shoesmith

The corrosion of nuclear fuel (UO2) in oxygenated solutions

Abstract The corrosion mechanism of UO2 (nuclear fuel) has been studied in 0.1 mol l−1 sodium perchlorate (pH = 9.5), with and without added sodium carbonate. The corrosion potential was followed for various exposure times. Subsequently, the electrode was either subjected to a cathodic stripping scan from the corrosion potential to −2.0 V to determine the presence and measure the thickness of surface films formed; or removed and examined by X-ray photo-electron spectroscopy to determine the composition of the electrode surface. In both perchlorate and perchlorate plus carbonate solutions two films were formed on the UO2 prior to the establishment of steady-state dissolution conditions. A layer of UO2.33 was formed over the first 10 h of exposure. The outer layers of this film slowly converted to hydrated UO3 (or uranyl carbonate) over the next ∼90 h. This conversion appeared to be concentrated at the grain boundaries. Corrosion rates were measured by extrapolating from the Tafel region for steady-state anodic dissolution to the corrosion potential. The corrosion process appears to be controlled by the kinetics of the anodic dissolution step.

Oxidation and dissolution of nuclear fuel (UO2) by the products of the alpha radiolysis of water

Oxidation of UO2 nuclear fuel by the products of the alpha radiolysis of water has been measured as a function of strength of the alpha flux and solution pH (0.1 mol L−1 NaClO4, 3.5≤pH≤11) using electrochemical techniques. Corrosion potentials were measured using a thin-layer corrosion cell in which an alpha source was brought within 30 μm of a UO2 electrode. Oxidative dissolution (corrosion) rates were then calculated as a function of alpha dose rate from the steady-state values of the corrosion potential using an electrochemical model. The corrosion rate was found to increase with an increase in alpha dose rate and with a decrease in pH for values <4. A procedure to predict the corrosion rate of used nuclear fuel in groundwater as a function of fuel cooling time is then described. As a consequence of the cell geometry used in corrosion potential measurements these predicted rates are appropriately applied to dissolution in cracks and fissures. The corrosion of fuel, supported solely by the alpha radiolysis of water, is predicted to be unimportant for CANDU reactor fuel with a burnup of 685 GJ/kg U for periods ≥600 a. However, for fuel with higher burnups, e.g., a typical PWR fuel (burnup 3888 GJ/kg U (45 MW d/kg U)), corrosion supported by the alpha radiolysis of water could be significant for time periods of ~2000 a. For periods greater than this (~600 a (CANDU); ~2000 a (PWR)) the oxidative dissolution can be appropriately considered as a chemical as opposed to corrosion reaction.

Anodic oxidation of copper in alkaline solutions: Part IV. Nature of the passivating film

Abstract Potentiostatic and X-ray photoelectron spectroscopic (XPS) techniques have been used to study the passivation of copper electrodes in 1.0 mol dm −3 LiOH. For potentials 2 O is present, and an upper layer of Cu(OH) 2 is formed by nucleation and growth from solution. The dissolved Cu 2+ ions necessary for Cu(OH) 2 precipitation are produced by metal dissolution in the pores of the Cu 2 O layer. Under these conditions, the surface is only partially passivated, since metal dissolution can continue in the pores of the base layer. For potentials >−60 mV (vs. SCE), these pores, and eventually the whole surface, are covered by a layer of CuO identified by XPS. When this layer is formed, the extent of Cu(OH) 2 formation is drastically reduced. In addition the cupric ion dissolution rate is reduced, indicating a much higher degree of surface passivation.

The effect of ph on the corrosion of nuclear fuel (UO2) in oxygenated solutions

Abstract The oxidative dissolution of UO2 has been studied in NaClO4 and Na2SO4 solutions as a function of pH over a range of 0.8 ⩽ pH ⩽ 12 using a combination of electrochemical and X-ray photoelectron spectroscopic techniques. The relative stability and solubility of solid uranium oxides and uranium speciation in aqueous solutions were examined using thermodynamic calculations. In neutral to alkaline solutions (pH ≥5), dissolution is preceded by the growth of a thin film of UO2 33 on the UO2 surface. This film achieves a steady-state thickness (∼6 nm) in 5 to 10 h, and the thickness increases with an increase in pH. Over the next 10 to 100 h, further oxidation to a hydrated form of UO3 occurs, after which steady-state dissolution conditions are achieved. Formation of UO3 · xH2O is a precursor to dissolution and its presence appears to be confined mainly to the grain boundaries. In acidic solutions UO2.33 formation does not occur. Oxidation proceeds directly to the UVI state (possibly to UO3 · xH2O) and the dissolution rate, measured by extrapolation of steady-state currents to the corrosion potential, is approximately 50 times greater at pH = 2.5 than it is at pH = 9.7. The electrochemically measured corrosion rate value of ∼45 × 10−8 g cm−2 d−1 at pH = 2.5 compares well with the value of ∼ 25 × 10−8 g cm−2d−1 measured chemically by Thomas and Till. The dependence of the steady-state corrosion potentials on pH suggests possible rate control by the anodic dissolution reaction for pH > 10. For pH values between 2 and 10, the corrosion potential varies only slightly with pH, and it is unclear whether the anodic or cathodic reaction is rate-controlling in this region.

Anodic oxidation of UO2: Part II. Electrochemical and X-ray photoelectron spectroscopic studies in alkaline solutions

Abstract The anodic oxidation of polycrystalline UO 2 has been studied in neutral Na 2 SO 4 solution (6 E E ≅+0.3 V), a thin film (t molecular layers) of UO 2+ x ( x ≤0.25) was formed. At slightly more positive potentials and longer times (1 t 3 O 7 grew on top of this UO 2+ x base layer. For potentials >+0.1 V, dissolution as U(VI) occurred, and for E =+0.3 V the surface film was transformed gradually to U 2 O 5 (≅1 h) and eventually to U 3 O 8 ( t t16 h). For E ≥+0.4 V a film of UO 3 · z H 2 O precipitated on the electrode surface in the absence of stirring. Also some dissolution of the U 2 O 5 film occurred due to a decrease in the local pH at the electrode surface. Both the precipitation and the film dissolution were prevented by stirring.

Hydrogen Absorption and the Lifetime Performance of Titanium Nuclear Waste Containers

For the conditions expected in a Canadian nuclear waste vault, the two corrosion processes most likely to lead to the failure of titanium waste containers are crevice corrosion and/or hydrogen induced cracking (HIC). In this report the processes likely to lead to hydrogen absorption by titanium alloys (Grades-2, -12, -16), and hence to render them susceptible to HIC are discussed. The possible paths to container failure via a combination of crevice corrosion, general passive corrosion and HIC are described and a criterion for container failure by HIC is defined. The modelling procedures developed to predict the consequences of hydrogen absorption are described, and the experimental methods used to measure required modelling parameters are discussed and the results summarized. The predictions of these models show that for alloys on which crevice corrosion either does not occur or is limited in extent (Ti-16 and Ti-12), container lifetimes of >10 a are achievable. As a result only those containers assumed to fail prematurely due to the presence of manufacturing defects need be included in assessment calculations.

Photothermal deflection spectroscopy investigations of uranium dioxide oxidation

Abstract Photothermal deflection spectroscopy (PDS) has been applied to the study of uranium oxide electrochemistry. PDS measures the optical absorption of the sample surface and concentration gradients formed in the electrolyte. Both these measurements are performed in situ under dynamic conditions. The combination of these two measurements provides information that can be used to infer the mechanism of the UO 2 surface chemistry. These studies of the uranium dissolution mechanism are performed in pH 10.5 sodium sulfate electrolytes at 22°C. The electrolytes are free from oxygen and complexing species. Our results suggest that dissolution of UO 2 can occur at oxidizing potentials as low as −300 mV/SCE. The optical absorption and concentration gradient results provide evidence for a surface recrystallization reaction that occurs at an oxidation potential of + 300 mV. The results show that the surface layer formed by the recrystallization reaction dissolves slowly by a non-electrochemical reaction.

Calculation of used nuclear fuel dissolution rates under anticipated Canadian waste vault conditions

Abstract Dissolution rates of UO2 fuel are determined as a function of alpha and gamma dose rates. These room-temperature rates are used to calculate the dissolution rates of used fuel at 100°C. Also, the alpha, beta and gamma dose rates in water in contact with the reference used fuel are calculated as a function of cooling time. These results are used to calculate used CANDU fuel dissolution rates as a function of time since emplacement in a defective copper container for the Canadian Nuclear Fuel Waste Management Program. It is shown that beta radiolysis of water is the main cause of oxidation of used CANDU fuel in a failed container and that the use of a corrosion model is required for ~1000 a of emplacement in the waste vault. The results obtained here can be adopted to calculate used nuclear fuel dissolution rates for other waste management programs.

Oxidation of Candu UO 2 Fuel by the Alpha-Radiolysis Products of Water

The oxidation of CANDU fuel (UO/sub 2/) by the alpha-radiolysis products of water has been investigated using electrochemical and X-ray photoelectron spectroscopic experiments. Experiments with O/sub 2/ and H/sub 2/O, two of the expected products of radiolysis of water, indicate that the rate of oxidation of UO/sub 2/ by H/sub 2/O/sub 2/ is about 200 times faster than by dissolved oxygen. Oxidation by both H/sub 2/O/sub 2/ and O/sub 2/ shows pH dependence. Possible reaction paths for the oxidation of UO/sub 2/ by radiolysis products are discussed.

Validation of an electrochemical model for the oxidative dissolution of used CANDU fuel

Abstract Measured dissolution rates of UO2 and used fuel powders are compared to dissolution rates predicted from electrochemical measurements on fuel pellets. This comparison was made for rates as a function of dissolved oxygen concentration, carbonate/bicarbonate concentration, and gamma irradiation dose rate. Measurements were also made as a function of temperature over the range 25–75°C. Good agreement was obtained between measured and predicted rates in aerated carbonate solutions confirming that the dissolution reaction is electrochemical in mechanism and that its rate can be predicted electrochemically. For non-complexing solutions agreement was not as good since electrochemical measurements on dissolution of fuel pellets appeared to be inhibited by the formation of secondary phases in occluded grain boundaries. Measured dissolution rates on powders do not appear to be affected in this manner since few occluded grain boundaries are present. Predicted and measured rates in gamma irradiated solutions, while measured under different conditions, showed the same general trends, and compare well to published literature values. In aerated non-complexing solutions and in aerated carbonate solutions the effect of gamma irradiation becomes insignificant below ∼1 Gy h−1 and ∼10 Gy h−1, respectively.