J.C. Wren

The Electrochemical Response of Preoxidized Copper in Aqueous Sulfide Solutions

The conversion of a Cu 2 O film on copper to Cu 2 S in aqueous sulfide solutions has been followed using a combination of electrochemical techniques and in situ Raman spectroscopy. Oxide films were electrochemically grown in alkaline solutions and their composition and morphology determined using Raman spectroscopy and scanning electron microscopy. Although corrosion potential measurements indicate that the aqueous sulfide solution rapidly penetrates the porous Cu 2 O layer, the oxide-to-sulfide reaction appears to proceed chemically at the oxide/solution interface rather than via the galvanic coupling of Cu oxidation to Cu 2 S and Cu 2 O reduction to Cu. In situ Raman spectroscopy confirms that the sulfide formed is Cu 2 S, and cathodic stripping voltammetry shows that the reaction is initially rapid and then proceeds at a constant rate until the conversion is complete. Comparison of the amounts of oxide initially present and sulfide eventually formed demonstrates that the conversion is 100% efficient. These studies are part of a larger project to determine the important corrosion processes on copper high-level nuclear waste containers exposed to anoxic aqueous sulfide containing groundwaters.

Corrosion Behavior of Uranium Dioxide in Alpha Radiolytically Decomposed Water

The response of the corrosion potential of a UO 2 electrode to oxidants (H 2 O 2 ,O 2 ) produced by the alpha radiolysis of water has been measured in a thin-layer electrochemical cell. This cell allows the electrode to be brought within ∼25 μm of a gold-plated alpha source, thereby ensuring the uniform distribution of radiolytic oxidants in the aqueous solution filling the gap. The kinetics of water radiolysis, coupled with the surface-catalyzed decomposition of H 2 O 2 and diffusive transport of radiolytic species out of the electrode-source gap, was modeled using a finite difference description of the cell and the commercial numerical integration software, Facsimile. Using this approach, the mechanism of UO 2 corrosion in the presence of alpha radiolysis was shown to be dominated by reaction of the fuel with radiolytically produced H 2 O 2 . To explain the results, two surface-catalyzed H 2 O 2 decomposition processes were invoked: one with H 2 to produce H 2 O, and a second process leading to the production of O 2 and H 2 O. For sufficiently high alpha source strengths, the fuel behavior becomes redox-buffered, i.e., independent of alpha source strength, due to the presence of a corrosion product deposit.

Iron oxyhydroxide colloid formation by gamma-radiolysis

Gamma-irradiation of deaerated aqueous solutions containing FeSO(4) leads to the formation of uniform-sized colloidal particles of γ-FeOOH. At short irradiation times, or in solutions with a low initial [Fe(2+)](0), spherical particles with a size less than 10 nm are formed. These primary particles grow to form a dendritic structure upon longer irradiation, and the final size of the large particles is ∼60 nm with a very narrow size distribution. Further prolonged irradiation does not change the final particle size. The narrow size distribution is attributed to rapid homogeneous radiolytic oxidation of soluble Fe(2+) to relatively insoluble Fe(3+) hydroxides [Fe(H(2)O)(6-n)(OH)(n)](3-n) leading to particle nucleation by spontaneous condensation. These primary particles then grow into γ-FeOOH particles with a dendritic structure. The final size reached at long times is regulated by the steady-state redox conditions established during long-term irradiation at the aqueous-solid interface.

Interaction of Aqueous Iodine Species with Ag2O ∕ Ag Surfaces

One of the safety issues of nuclear power plants is the potential radiation dose to the public in the unlikely event of a severe accident. From the perspective of public safety, radioiodine is one of the most important fission products from the uranium fuel because of its large fuel inventory, high volatility, and radiological hazard. Such an incident would lead to fuel and fuel channel damage, and it is assumed that a significant fraction of the radioiodine fuel inventory would be released from the reactor core into the containment building. 1-3 It has been established that most of the released iodine would quickly dissolve and remain in the water ubiquitous throughout the containment building following an accident. 4 However, a small fraction could be released to the gas phase due to the continuous conversion of nonvolatile to volatile iodine species under the highly oxidizing conditions prevailing in the presence of ionizing radiation. 5 Because of its mobility, the gaseous iodine concentration is a critical parameter for safety assessment and postaccident management. One reaction of interest in assessing iodine volatility is that between aqueous iodine species and silver surfaces. The control rod for neutron flux in some nuclear reactors is made of silvercadmium, and silver is assumed to be released from the reactor core into the containment building environment in some accident scenarios, and its reaction with aqueous iodine to form insoluble silver iodide could dominate iodine behavior. In fact, a significant reduction in iodine volatility in the presence of silver has been observed in engineering-scale experiments simulating postaccident containment conditions. 6,7 In the highly oxidizing and ionizing radiation conditions expected, Ag could enter the aqueous phase in either the metallic or oxidized AgOH, Ag2O, and AgO forms. The availability of aque

Modelling development of acidification within corroding sites on spent fuel surfaces

Abstract A model has been developed to predict whether the development of acidity is feasible within actively corroding sites on spent nuclear fuel (UO2) surfaces inside a failed nuclear waste container. The model simulations demonstrate that the build-up of acidity is possible within flaws and pores in a corroded UO2 surface, providing the separation of anodes and cathodes occurs. The extent to which the pH can be depressed is determined by the dissolution rate of the fuel, the dimensions of the defect, the local redox conditions which determine the corrosion potential, and the fraction of the fuel surface that is reactive. Based on the anticipated redox conditions established radiolytically in a failed container it is shown that a suppression of the pH sufficient to accelerate dissolution (pH≤5) is very unlikely.

Modeling the Distribution of Acidity within Nuclear Fuel (UO 2 ) Corrosion Product Deposits and Porous Sites

A model for acidity within pores within corrosion products on anodically-dissolving UO{sub 2} was developed using Comsol Multiphysics 3.2 to complement ongoing electrochemical measurements. It was determined that a depression of pH within pores can be maintained if: electrochemically measured dissolution currents used in the calculations are attenuated to reflect very localized pores; corrosion potentials exceed -250 mV (vs. SCE); and pore depths are >1 {mu}m for 300 mV or >100 {mu}m for -50 mV (vs. SCE). Mixed diffusional-chemical equilibria control is suggested through deviations in the shapes between pH-potential and pH-pore depth plots. (authors)