Nuclear Waste Disposal Conditions Research Articles

The Influences of Traces of Oxygen and Sulfide on the Corrosion of Copper Under Canadian Nuclear Waste Disposal Conditions

The Canadian repository concept for the permanent disposal of used nuclear fuel is to seal it in Cu-coated steel containers and dispose of it in a deep geological repository (DGR) surrounded by compacted bentonite clay. The Cu coating will be directly bonded on the steel vessel and lid using electro-deposition (ED-Cu), with the final container closure weld sealed with a cold-sprayed Cu deposit (CS-Cu). To assess the durability of such a container, long-term corrosion experiments have been performed on CS-Cu and ED-Cu specimens, and compared with those performed on wrought Cu specimens provided by the Swedish Nuclear Fuel and Waste Management Company (SKB-Cu).Experiments were conducted in 3 M NaCl solutions containing traces of sulfide (< 10- 6 M) for exposure periods of ≥150 days. These conditions reflect a generic groundwater that may influence container corrosion in a Canadian DGR Since conditions in a DGR will evolve from initially oxic to eventually anoxic, experiments were conducted under three sets of redox conditions: (i) aerated (oxic) conditions; (ii) Ar-purged (oxic to anoxic transition conditions (with a dissolved [O2] ~ 10- 6 M)); and (iii) anaerobic chamber (anoxic) conditions ([O2] < 1.3 × 10- 9 M). The exposure process was followed using open circuit (i.e., corrosion) potential and polarization resistance (RP) measurements. Specimens were analyzed before and after corrosion using optical microscopy, scanning electron microscopy on surfaces and focused ion beam cut cross sections, Raman spectroscopy, and X-ray photoelectron/Auger spectroscopy.In all environments, all three types of specimens (ED-Cu, CS-Cu, SKB-Cu) showed similar behaviour. Under anoxic conditions, minor corrosion, indicated by large RP values, leading to the deposition of traces of chalcocite (Cu2S) (identified by Raman spectroscopy) and observed by SEM, indicated that corrosion could be attributed to reaction with the trace amounts of sulfide present in NaCl. In Ar-purged solutions, the RP values were approximately one order of magnitude lower (i.e., the corrosion rate was one order of magnitude higher) than those measured under anoxic conditions, and significant corrosion damage was observed in SEM micrographs. Both Cu2O and Cu2S were detected by Raman spectroscopy. The detection of oxidized sulfur species (SO3 2 - and SO4 2 -) on the corroded surface by XPS/Auger confirmed that even trace amounts of dissolved O2 led to sulfide oxidation and the possible formation of potentially more corrosive species such as thiosulfate (S2O3 2 -). Under aerated conditions, extensive corrosion occurred leading to the formation of green CuII products (atacamite (Cu2(OH)3Cl)) with no detectable Cu2S formation.Within the DGR, corrosion of Cu under anoxic conditions will be limited by the available supply of sulfide, which may be produced remotely at the host rock/clay interface and/or in the porewater within the rock fractures. In the presence of even trace of dissolved O2, the corrosion process is significantly accelerated with the formation of potentially aggressive oxidized sulfur species. Under aerated conditions, extensive corrosion due to reaction with O2 leads to the deposition of CuII deposits. On the time scale of these experiments with only trace of sulfide present, no detectable conversion of copper oxide to Cu2S was observed. Keywords: Copper; Sulfide; Corrosion; Nuclear waste disposal

The influence of hydrogen permeation on the protection performance of the Cu coating of nuclear waste containers

Hydrogen permeation and corrosion behaviour of Cu-coated carbon steel under simulated nuclear waste disposal conditions of the Beishan area have been investigated. The permeation current (with −500 μA·cm−2 charging current) decreased from 3778 to 141 and 47 nA.cm−2 respectively when 5 and 10 µm thick Cu coatings were electrodeposited on the steel. The XRD analysis suggests the formation of CuH during the hydrogen charging treatment, and its subsequent decomposition formed a protective CuOH/Cu2O layer on the charging surface when exposed to simulated groundwater for 44 h. No damage to the Cu coating-steel interface was observed in the cross-section analysis.

Predicting zeolites’ stability during the corrosion of nuclear waste immobilization glasses: Comparison with glass corrosion experiments

During the long-term corrosion of nuclear waste glasses under nuclear waste disposal conditions, the precipitation of zeolitic phases has been linked to a delayed acceleration in glass corrosion (known as “Stage III”). Hence, predicting the thermodynamic propensity for zeolites to form upon the dissolution of nuclear waste glasses is key to ensure their long-term performance. Here, we compile a unified, internally-consistent thermodynamic database “clay20” to estimate the stability of clay and feldspar phases relevant to nuclear waste immobilization glasses, including beidellite(Mg, Ca, Na, K), kaolinite, montmorillonite(Mg, Ca, Na, K), nontronite(Mg, Ca, Na, K), saponite(Ca, Na, K), and albite. Based on this, we report a geochemical modeling method allowing us to predict the stability of secondary phases (including zeolites, calcium–silicate–hydrate gels, and clays) upon the dissolution of nuclear waste immobilization glasses. We show that this approach offers a realistic description of the stability of the secondary phases forming during the dissolution of two archetypical model nuclear glasses (namely, the International Simple Glass, ISG, and WVUTh-203) under conditions relevant to nuclear waste disposal (T = 90 °C, p = 1 bar) as a function of pH. We find that the formation of silica and clay secondary phases is thermodynamically favored at low pH (pH < 10), whereas, in contrast, zeolite (analcime) and calcium–silicate–hydrate phases are favored at high pH (pH > 10.5). This suggests that thermodynamics (i.e., not solely kinetics) might play a key role in determining the range of solution pH wherein stage III corrosion may occur, i.e., when zeolite formation is favored.

Corrosion behavior of aluminum alloy 5754 in cement‐based matrix‐simulating nuclear waste disposal conditions

AbstractDepending on the lifetime and level of radioactivity of radioactive wastes, different disposal facilities are considered. Though low‐ and intermediate‐level short‐lived waste can be disposed in surface disposal facilities, deep geological disposal is considered for high‐ and intermediate‐level long‐lived waste. In France and Belgium, long‐term disposal is studied in clay host rock media. For aluminum, the disposal concept is based on encapsulation of the waste in a cement‐based matrix. It is also well‐known that aluminum is prone to severe corrosion in sufficiently alkaline environments leading to possible hydrogen production. To ensure the safety of the disposal facilities and the integrity of the cement capsules, the amount of aluminum that is disposed in each waste package must be specified and is limited to mitigate the level of hydrogen production by aluminum corrosion. In the present study, the corrosion resistance of an aluminum alloy (grade EN‐AW‐5754/H111) in two different cement matrices was studied in different configurations at room temperature. In each case, the evolution of hydrogen production was monitored to address the corrosion rate variation versus time.

Technetium Incorporation into C14 and C15 Laves Intermetallic Phases

ABSTRACTLaves-type intermetallic phases have been observed to be the dominant phases in a series of alloy compositions being designed for the immobilization of technetium in a metallic waste form. The dominant metals in the alloy compositions were Fe-Mo and Fe-Mo-Zr. The alloy composition, Fe-Mo-Zr, also contained Pd, Zr, Cr, and Ni. Both non-radioactive rhenium-containing and radioactive technetium-bearing alloy compositions were investigated. In the Fe-Mo series, the phases observed were Fe2Mo (C14 Laves phase) and ferrite in agreement with predictions. Both Tc and Re resided predominantly in the Laves phases. In the Fe-Mo-Zr system, the phases included hexagonal C14 with the composition (Fe,Cr)2Mo, cubic C15 phase with a (Fe,Ni)2Zr composition, and the hcp phase Pd2Zr. The observation of these phases was in agreement with predictions. Re was found in the C14 intermetallic, (Fe,Cr)2Mo. Technetium was also observed to be partitioned preferentially into the (Fe,Cr)2Mo phase; however, this phase exhibited a cubic structure consistent with the C15 structural type. The composition of Laves phases is influenced by both the atomic size and electro-negativity of the constituent elements. The long-term release behavior of technetium under nuclear waste disposal conditions may be more dependent on the corrosion characteristics of these individual Laves phases containing Tc than the other metallic phases.

Copper Corrosion in Aqueous Sulfide Solutions under Nuclear Waste Repository Conditions

ABSTRACTThe corrosion behavior of oxygen-free copper in anoxic sulfide solutions under nuclear waste disposal conditions was studied using electrochemical impedance spectroscopy (EIS), scanning electron microscopy (SEM) equipped with a focused ion beam (FIB), X-ray photoelectron spectroscopy (XPS) and Micro X-ray diffraction (μXRD). The film growth process and mechanism were elucidated using an Au marker test, and the contribution of solution diffusion to corrosion was demonstrated in magnetically-stirred experiments. The effect of groundwater chemistry, particularly chloride content on copper corrosion and film properties was characterized using long-term corrosion experiments.

Kinetics of pyrite to pyrrhotite reduction by hydrogen in calcite buffered solutions between 90 and 180°C: Implications for nuclear waste disposal

Abstract The kinetics of abiotic redox reactions induced by hydrogen are poorly documented although it represents a growing area of interest in terms of both nuclear waste storage assessment and the comprehensive study of hydrogen-rich fluid in mid-ocean ridge hydrothermal systems. We present an experimental kinetics study of pyrite reduction into pyrrhotite under significant H2 pressure and mid-hydrothermal conditions. We describe the mechanism and kinetic behavior of this reaction by combining textural and solution analyses under various conditions of temperature, pyrite particles size, H2 pressure and pH. When pH is controlled by calcite, the reaction presents all the characteristics of a coupled dissolution–precipitation mechanism occurring at the pyrite–pyrrhotite interface. By considering the chemical affinity of the coupled reaction as a function of reaction extent, we demonstrate that the spatial coupling is induced both by pyrite as a substrate for pyrrhotite nucleation and by the role of fluid chemistry at the reaction front. Far from equilibrium with respect to pyrite, the kinetics of sulfide production associated with the reaction are linearly related to the square root of time with an activation energy of 53 kJ/mol. This value is higher than what is expected for a diffusion-controlled kinetic regime. We suggest that the reaction rate is controlled both by pyrite reductive dissolution and by sulfide diffusion through the porous pyrrhotite microstructure. We provide a simple sulfide production-rate expression on the basis of our measured rate constants that can be used in geochemical modeling to further evaluate the impact of hydrogen on pyrite under nuclear waste disposal conditions.

Electrochemical, SECM, and XPS Studies of the Influence of H2 on UO2 Nuclear Fuel Corrosion

ABSTRACTA combination of electrochemical, scanning electrochemical microscopy (SECM), X-ray photoelectron spectroscopic (XPS), and time of flight secondary ion mass spectrometry (TOF - SIMS) techniques are being employed to investigate the effects of dissolved hydrogen on the aqueous corrosion of uranium dioxide (UO2) under nuclear waste disposal conditions. Corrosion potential (ECORR) measurements indicate that the oxidation of dissolved hydrogen on noble metal ε-particles polarizes the UO2 nuclear fuel surface to reducing potentials (−300 – −400 mV vs. SCE; i.e., to ECORR values more negative than those observed under anoxic (argon-purged) conditions (−200 mV vs. SCE). A comparison of the behaviors observed on SIMFUEL specimens with and without incorporated noble metal ε-particles indicate that these particles may act as catalytic electrodes for H2 oxidation. It is the galvanic coupling of these particles to the UO2 matrix which suppresses the fuel corrosion potential thereby preventing oxidation of the fuel surface.

The inhibiting effects of hydrogen on the corrosion of uranium dioxide under nuclear waste disposal conditions

A number of electrochemical experiments were employed to investigate the effects of hydrogen on the corrosion of UO 2 under nuclear waste disposal conditions. A combination of corrosion potential ( E CORR) measurements and cyclic voltammetry have indicated that dissolved hydrogen can polarize the UO 2 surface to reducing potentials; i.e., to E CORR values more negative then those observed under anoxic (argon-purged) conditions. A comparison of the behaviours of SIMFUEL specimens with and without incorporated noble metal ε-particles indicates that these particles may act as catalytic electrodes for H 2 oxidation, H 2 ↔ 2 e − + 2H +. It is the galvanic coupling of these particles to the UO 2 matrix which suppresses the fuel corrosion potential.

Application of electrochemical methods in the development of models for fuel dissolution and container corrosion under nuclear waste disposal conditions

The permanent disposal of nuclear fuel wastes requires the development of models that can assess the performance of a disposal vault over long periods of time. Models to assess the long-term stability of the nuclear fuel (UO2) and the corrosion performance of the waste container (either copper or titanium) have been based on electrochemical principles. Here we review the chemical/electrochemical performance of fuel and the two candidate container materials, and describe some of the electrochemical studies undertaken either to develop the mechanistic understanding upon which these models are based or to measure the values of parameters required to evaluate long-term performance. These include the following: the anodic dissolution of UO2; the reduction of O2 on various specimens of UO2; the crevice corrosion of various titanium alloys; the impedance characteristics of passive films on Ti alloys; the anodic dissolution of copper in chloride solutions; the reduction of O2 on copper; the effect of various transport barriers on the corrosion of copper; and the prediction of the corrosion potential of copper in aerated chloride solutions. Keywords: uranium dioxide, copper, titanium, nuclear waste, oxygen.

Partitioning of La between solid and solution during the ageing of Si [sbnd]Al [sbnd]Fe [sbnd]La [sbnd]Ca gels under simulated near-field conditions of nuclear waste disposal

Lanthanum-bearing gels of various stoichiometries (Si Fe ± Al ± Ca) and structures (hydrated Na-bearing and anhydrous Na-free materials), were aged in oxidizing aqueous solutions at 90°C and under a strong thermal gradient. The presence of La in the starting material did not inhibit the crystallization of ferrous clays (nontronite) for compositions with ratios Si/(Fe + Al) ⩽ 2, whereas crystallization failed forSi(Fe + Al) > 2.7. Clays are able to trap La, probably as an interlayer cation. In some experiments, a significant amount of La was also associated with newly formed silicated amorphous phases (APs), the precipitation of which from the solution was enhanced by a decrease of temperature. Extrapolation of these results to nuclear waste disposal conditions suggests that the radionuclide-rich hydrated amorphous layer (AL) which forms during the leaching of nuclear waste glasses is likely not to crystallize into clays. The hydrated amorphous layer should thus be a stable secondary barrier against the dispersion of radionuclides from leached nuclear waste glasses. Towards the far-field conditions, heavy elements are likely to be found in newly formed clays and silicated amorphous phases.

Microbial Effects on the Radionuclide Transport in a Deep Nuclear Waste Repository

ABSTRACTThis study deals with the effects of microorganisms on the transport of radionuclides under deep nuclear-waste disposal conditions. Metabolism of a cellulolytic microorganism is studied. Cellulose, as a carbon source, is representative of nuclear waste.A pilot device allows the study of the general effect of microrganisms. Bioleaching of radionuclides by a fungal culture is performed on columns of clay and cement used as engineered barriers. Cesium and Uranium had been incorporated into matrices prior to the tests. Operating conditions have been choosen according to realistic conditions of a deep repository. The production of organic acids by microorganisms is qualitatively and quantitatively determined.In addition, direct effects of microorganisms (biosorption, bioaccumulation) and indirect effects (complexing agents issued from the diodegradation of cellulose) on the transport and/or retardation of radionuclides are studied.

Interaction of Water and Compacted Sodium-Bentonite in Simulated Nuclear Waste Disposal Conditions

ABSTRACTAlteration of compacted sodium-bentonite (Volclay MX-80) caused by groundwater in simulated repository conditions for high level radioactive waste, was studied in an experiment where bentonite (wrapped by a copper cylinder) was let to react with two types (A,B) of synthetic granitic groundwater that are distinguished by their initial concentration of potassium and chloride. The reaction took place in ambient conditions at a temperature of 75 °C and proceeded during several time intervals up to 36 months.At the end of each time interval the water was chemically analysed for determination of possible changes in composition. Chemical and mineralogical changes in the bentonite were investigated by using NH4Cl extractions, XRD and microprobe (SEM, EDS) analyses and were studied as a function of the reaction time (months) as well as of the distance (mm) from the contact front with water.Sodium ions were found to migrate out from the bentonite while being replaced by other cations such as calcium, magnesium and to some extent, particularly during the reaction of the bentonite with water B, by potassium. No clear evidence was found for the fixation of potassium ions in the interlayer position of montmorillonite clay and the transformation to illite. The main mineralogical change in the bentonite was from Na- to Ca-rich montmorillonite. Secondary processes were the dissolution-precipitation of sulphur compounds, dissolution of gypsum and carbonates and the dissolution-precipitation of copper compounds.

Statistical study of pit propagation in carbon steel under nuclear waste disposal conditions

AbstractOne aspect of a project aimed at specifying the metal allowance required to prevent the corrosion penetration of carbon steel nuclear waste containers over a 1000 year period in a geological repository is described. This involves a statistical analysis of long term pit growth data from potentiostatically controlled tests in 0·1 M NaHCO3 + 1000 ppm Cl−solution at 90°C. In general it was found that these data fitted unlimited exponential distribution functions. This was surprising because it was anticipated that charge and mass transport effects would set an upper limit to pit depth. That this limit was not detected may be because it is large, or because the specimen surface area was insufficient to detect the ‘tail off” characteristic of a limited distribution. Using the exponential distributions it has been shown that for probabilities less than 0·2 the pit depth x(mm) having a probability PA(x) of occurring in a surface area A increases with time t(h) according to the expression: x = 9·5 × 10−5 t...