The effect of acetohydroxamic acid on stainless steel corrosion in nitric acid

We present the first study of the effect of acetohydroxamic acid (AHA) on the corrosion behaviour of stainless steels. Particularly, studies have been performed using steels and physico-chemical conditions equivalent to those proposed for use in advanced nuclear reprocessing platforms. In these, AHA has been shown to have little effect on either steel passivation or reductive dissolution of both SS304L and SS316L. However, under transpassive dissolution conditions, AHA while in part electrochemically oxidised to acetic acid and nitroxyl/hydroxylamine, also complexes with Fe3 +, inhibiting secondary passivation and driving transpassive dissolution of both steels

The effect of SO3-Ph-BTBP on stainless steel corrosion in nitric acid

SO3-Ph-BTBP is a hydrophilic tetra-N-dentate ligand proposed for An(III)/Ln(III) separation by solvent extraction, and a candidate for use in future advanced reprocessing schemes such as GANEX and SANEX. We present the first study of the effect of SO3-Ph-BTBP on the corrosion behavior of stainless steels. Specifically, studies have been performed using steels and conditions equivalent to those found in relevant nuclear reprocessing flow sheets. SO3-Ph-BTBP has been shown to have little effect on either steel passivation or reductive dissolution. However, if driven cathodically into a region of hydrogen evolution at the electrode surface or conversely anodically into a region of transpassive dissolution, observed currents are reduced in the presence of SO3-Ph-BTBP, suggesting corrosion inhibition of the steel potentially through weak absorption of a SO3-Ph-BTBP layer at the metal-solution interface. The lack of any observed corrosion acceleration via complexation of Fe3+ is surprising and has been suggested to be due to the slow extraction kinetics of SO3-Ph-BTBP as a result of a requirement for a trans- to cis-conformational change before binding

Nitric acid reduction on 316L stainless steel under conditions representative of reprocessing

Steels comprise the largest class of metal-based materials encountered on nuclear sites. An understanding of how process steels interact with HNO3 in spent fuel treatment plant environments is required to enable informed decisions to be made about the design and effective application of different steel types within nuclear environments. Stainless steels readily passivate in nitric acid. However, increasing the oxidising power of the media can lead to passive film dissolution, resulting in rapid transpassive corrosion. The corrosion of steels in nitric acid is further complicated by the autocatalytic reduction of HNO3 to aqueous HNO2 which attacks the steel surface. This paper describes the effect of this behaviour on process steels in stagnant and/or flowing conditions using electrochemical and microgravimetric based methods. We describe linear sweep voltammetry studies performed on a 316L stainless steel rotating disk electrodes in varying concentrations of nitric acid and rotation speeds and provide a qualitative interpretation of the results and what these imply about the mechanism of HNO3 reduction. These findings will be used in follow on studies to determine the kinetic parameters of the nitric acid reduction reaction at the surface of 316L stainless steel

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

For 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 H2O2-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 [H2O2] 0.1 mol dm(-3) 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 UO2 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 [H2O2] 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. (C) 2015 Elsevier B.V. All rights reserved

Photocatalytically Driven Dissolution of Macroscopic Nickel Surfaces

Photocatalytically generated H2O2-driven nickel dissolution has been studied as a novel, secondary waste minimal decontamination process for nuclear process steels. Nickel corrosion experiments in dilute H2SO4 show that at deliberately added [H2O2] ≤ 1 mM, nickel dissolution occurs via formation and dissolution of NiOH groups; at [H2O2] ≥ 10 mM (pseudo-)passivation by NiO prevents this. Furthermore, Nickel also dissolves slowly in mild acid, dissolution that is significantly accelerated in the presence of photogenerated peroxide – suggesting that photocatalytically generated H2O2 could be used to selectively increase dissolution of Ni, and potentially steel, surfaces that normally dissolve only slowly in mild acid

Corrosion behaviour of AGR simulated fuels:evolution of the fuel surface

We have prepared a range of Advanced Gas-cooled Reactor (AGR) SIMFUELs at a range of simulated burn-ups and, using Raman spectroscopy, have studied the effect of the SIMFUEL dopants on the UO2 crystal structure. We have also studied the effect of exposure to hydrogen peroxide solutions on the SIMFUEL surface. The intensity of the fundamental U-O stretch (445 cm-1) decreases as the amount of dopant increases in each SIMFUEL burn-up composition. A simultaneous increase in the lattice damage (500 – 700 cm-1) peak is observed as the UO2 cubic fluorite lattice structure becomes more distressed and moves towards a tetragonal structure. Exposure to 100 µmol dm-3 H2O2 further decreases the fundamental U-O stretch and increases the lattice damage peak, suggesting that additional point defects are established as the concentration of interstitial oxygen is increased in the lattice via the H2O2-induced corrosion of the SIMFUEL

Direct mass analysis of water absorption onto ceria thin films

Plutonium oxide (PuO2) is one of the most highly radioactive components of nuclear fuel waste streams and its storage poses particular challenges due to the high temperatures produced by its decay and the production of gases (particularly H2 and steam). Its high radiotoxicity necessitates the use of analogues, such as CeO2, to allow the comprehensive study of its interaction with water under storage conditions. We have developed a method which enables direct gravimetric measurement of water adsorption onto CeO2 thin films with masses in the microgram region. Porous CeO2 films were fabricated from a surfactant based precursor solution. The absorption of water onto the CeO2 coating at different relative humidities was studied in a closed reactor. Quartz Crystal Microbalance (QCM) gravimetry was used as a signal transducer, as changes in crystal resonant frequency due to absorbed mass are directly and linearly related to mass changes occurring at the crystal surface. Using this method, we have determined the enthalpy of absorption of water onto CeO2 to be 49.7 kJmol–1 at 75°C, 11 kJmol-1 greater than the enthalpy of evaporation. This enthalpy is within the range predicted for the absorption of water onto PuO2, indicating this method allows for investigation of water absorption using microgram samples

Raman Studies of Advanced Gas-Cooled Reactor Simulated Spent Nuclear Fuels

Analysis of advanced gas-cooled reactor (AGR) simulated used nuclear fuels (SIMFuels) has been carried out using micro-Raman spectroscopy in order to understand the effect lanthanide species (e.g. Nd, Y, Ce), representative of fission products generated during fuel burnup, have on the structure of the UO2 matrix in spent AGR fuel. Results show a decrease in perfect fluorite character with increasing burnup as well as the development of a broad lattice distortion peak between 500 and 650 cm-1. Peak analysis of this broad band reveals in it comprised of three overlapping peaks at 534 cm-1, 574 cm-1 and 624 cm-1. The peak at 534 cm-1 has been examined and is suggested to be due to a local phonon mode associated with oxygen-vacancy-induced lattice distortion as a result of lanthanide 3+ ion incorporation into the UO2 bulk matrix

Mechanical Properties of Advanced Gas-Cooled Reactor Stainless Steel Cladding After Irradiation

The production of helium bubbles in advanced gas-cooled reactor (AGR) cladding could represent a significant hazard for both the mechanical stability and long-term storage of such materials. However, the high radioactivity of AGR cladding after operation presents a significant barrier to the scientific study of the mechanical properties of helium incorporation, said cladding typically being analyzed in industrial hot cells. An alternative non-active approach is to implant He2+ into unused AGR cladding material via an accelerator. Here, a feasibility study of such a process, using sequential implantations of helium in AGR cladding steel with decreasing energy is carried out to mimic the buildup of He (e.g., 50 appm) that would occur for in-reactor AGR clad in layers of the order of 10 lm in depth, is described. The implanted sample is subsequently analyzed by scanning electron microscopy, nanoindentation, atomic force and ultrasonic force microscopies. As expected, the irradiated zones were affected by implantation damage (<1 dpa). Nonetheless, such zones undergo only nanoscopic swelling and a small hardness increase (10%), with no appreciable decrease in fracture strength. Thus, for this fluence and applied conditions, the integrity of the steel cladding is retained despite He2+ implantation

Fixed Contamination on Steel Surfaces: First Use of Quartz Crystal Microgravimetry to Measure Oxide Growth on Process Steels Under Conditions Typical of Nuclear Reprocessing

During the lifetime of a nuclear facility, radioactive material may become deposited onto process and structural material surfaces. Due to their high corrosion resistance, steels comprise the largest class of metal-based materials encountered on nuclear sites. A greater understanding of the mechanisms of how contaminant radionuclides interact with and attach to process steels in nuclear plant environments is required in order to enable informed decisions to be made about the design and effective application of decontamination techniques, reducing secondary wastes. There is limited literature relating to radionuclide sorption mechanisms on steels. Key studies have found that sorbed contamination is almost entirely located in the outermost oxide layers formed at steel surfaces. Thus, a molecular level investigation of contaminant uptake during induced oxide formation would be beneficial in developing steel decontamination strategies. Stainless steel 316L is commonly employed in the nuclear industry in process streams and pipework. Thus, we describe work carried out on electrochemically accelerated oxide growth on 316L and SS2343 (a 316L analog) in nitric acid media and its characterisation using combined voltammetric and microgravimetric measurements. These allow identification of active, passive, high voltage passive, transpassive and secondary passivation regimes in the associated current voltage curves. EQCM on SS2343 coated quartz crystal piezoelectrodes, combined with potentiodynamic polarisation data have allowed us to determine that fastest net growth of surface oxide occurs in the low voltage passive regime. Further, we have directly measured the growth of that layer by using in situ microgravimetry for the first time. We will be shortly using the methods described above and radionuclide surrogates for the study of contaminant uptake during oxide formation and uptake onto preformed oxide layers. XPS will be used to determine layer composition and mode of contaminant uptake