The effect of hydrogen peroxide on uranium oxide films on 316L stainless steel

Abstract

h i g h l i g h t sThe first report of the presence of both UO 2 and polymeric UO 2 2+ in the same electrodeposited U oxide sample.The action of H 2 O 2 on electrodeposited U oxides is described using corrosion based concepts.Electrodeposited U oxide freely dissolves at hydrogen peroxide concentrations <100 lmol dm À3 .At [H 2 O 2 ] > 0.1 mmol dm À3 dissolution is inhibited by formation of a studtite passivation layer. At [H 2 O 2 ] P 1 mol dm À3 studtite formation competes with uranyl-peroxide complex formation. a b s t r a c tFor the first time the effect of hydrogen peroxide on the dissolution of electrodeposited uranium oxide films on 316L stainless steel planchets (acting as simulant uranium-contaminated metal surfaces) has been studied. Analysis of the H 2 O 2 -mediated film dissolution processes via open circuit potentiometry, alpha counting and SEM/EDX imaging has shown that in near-neutral solutions of pH 6.1 and at [H 2 O 2 ] 6 100 lmol dm À3 the electrodeposited uranium oxide layer is freely dissolving, the associated rate of film dissolution being significantly increased over leaching of similar films in pH 6.1 peroxide-free water. At H 2 O 2 concentrations between 1 mmol dm À3 and 0.1 mol dm À3 , formation of an insoluble studtite product layer occurs at the surface of the uranium oxide film. In analogy to corrosion processes on common metal substrates such as steel, the studtite layer effectively passivates the underlying uranium oxide layer against subsequent dissolution. Finally, at [H 2 O 2 ] > 0.1 mol dm À3 the uranium oxide film, again in analogy to common corrosion processes, behaves as if in a transpassive state and begins to dissolve. This transition from passive to transpassive behaviour in the effect of peroxide concentration on UO 2 films has not hitherto been observed or explored, either in terms of corrosion processes or otherwise. Through consideration of thermodynamic solubility product and complex formation constant data, we attribute the transition to the formation of soluble uranyl-peroxide complexes under mildly alkaline, high [H 2 O 2 ] conditions -a conclusion that has implications for the design of both acid minimal, metal ion oxidant-free decontamination strategies with low secondary waste arisings, and single step processes for spent nuclear fuel dissolution such as the Carbonate-based Oxidative Leaching (COL) process.

Cesium and Strontium Contamination of Nuclear Plant Stainless Steel: Implications for Decommissioning and Waste Minimization

“…5 The selection of stainless steel in the nuclear industry is based on a combination of high radiation stability and excellent corrosion resistance, the latter afforded by the spontaneous formation of a nanometer-scale passivating Cr-oxide layer at the steel surface. 6 It has been assumed that deposited radioactivity becomes bound to/impregnated within this passive layer during the accumulation process, 7,8 although these phenomena have not been fully demonstrated. However, to make informed decisions regarding the design and optimization of effective decontamination treatments for these materials, an understanding of the principal chemical interactions driving radionuclide uptake onto or into the steel is critical.…”

Efficient electrosorption of uranyl ions by a homemade amidoxime‐modified carbon paper‐based electrode in acidic aqueous condition

“…UO 2 2+ could be reduced to lower valence (U 4+ ) at a certain potential 53,54 . It meant that UO 2 2+ was reduced to UO 2 at the negative electrode, then reacted with H 2 O 2 (reduction of the dissolved oxygen) on the negative electrode to form (UO 2 )O 2 ·4H 2 O 55,56 . During CP‐AO electrosorption, UO 2 2+ was electrochemically reduced to UO 2 after it combined with the amidoxime group, then UO 2 reacted with dissolved oxygen in the solution and oxygen generated by the anode reaction to form UO 3 ·2H 2 O 57 .…”

Impact of Stoichiometry on the Mechanism and Kinetics of Oxidative Dissolution of UO2 Induced by H2O2 and γ-Irradiation

“…This process has been studied extensively over many decades and it has been found that H 2 O 2 is the radiolytic oxidant that dominates the oxidation process under conditions relevant for the safety assessment of a deep geological repository for the spent nuclear fuel. Oxidation of UO 2 by H 2 O 2 is rather complicated mainly for two reasons: − (1) the oxidation process contains multiple elementary reaction steps, and (2) the redox property of the UO 2 matrix is strongly dependent on the composition.…”