Nuclear Waste Containers Research Articles

Characterization and modeling of major constituent equilibrium chemistry of a blended cement mortar

Cementitious materials containing ground granulated iron blast furnace slag and coal combustion fly ash as admixtures are being used extensively for nuclear waste containment applications. Whereas the solid phases of ordinary Portland cement (OPC) have been studied in great detail, the chemistry of cement, fly ash and slag blends has received relatively less study. Given that OPC is generally more reactive than slag and fly ash, the mineralogy of OPC provides a logical starting point for describing the major constituent chemistry of blended cement mortars. To this end, a blended cement mortar containing Portland cement, granulated blast furnace slag, fly ash and quartz sand was modeled using a set of solid phases known to form in hydrated OPC with the geochemical speciation solver LeachXS/ORCHESTRA. Comparison of modeling results to the experimentally determined pH-dependent batch leaching concentrations (USEPA Method 1313) indicates that major constituent concentrations are described reasonably well with the Portland cement mineral set; however, modeled and measured aluminum concentrations differ greatly. Scanning electron microscopic analysis of the mortar reveals the presence of Al-rich phyllosilicate minerals heretofore unreported in similar cementitious blends: kaolinite and potassic phyllosilicates similar in composition to illite and muscovite. Whereas the potassic phyllosilicates are present in the quartz sand aggregate, the formation of kaolinite appears to be authigenic. The inclusion of kaolinite in speciation modeling provides a substantially improved description of the release of Al and therefore, suggests that the behavior of phyllosilicate phases may be important for predicting long-term physico-chemical behavior of such systems.

Corrosion susceptibility of steel drums to be used as containers for intermediate level nuclear waste

The present work is a study of the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins contaminated with different types and concentrations of aggressive species. A special type of specimen was manufactured to simulate the cemented ion-exchange resins in the drum. The evolution of the corrosion potential and the corrosion rate of the steel, as well as the electrical resistivity of the matrix were monitored over a time period of 900 days. The aggressive species studied were chloride ions (the main ionic species of concern) and sulphate ions (produced during radiolysis of the cationic exchange-resins after cementation). The work was complemented with an analysis of the corrosion products formed on the steel in each condition, as well as the morphology of the corrosion products. When applying the results obtained in the present work to estimate the corrosion depth of the steel drumscontaining the cemented radioactive waste after a period of 300 years (foreseen durability of the Intermediate Level Radioactive Waste facility in Argentina) , it is found that in the most unfavourable case (high chloride contamination), the corrosion penetration will be considerably lower than the thickness of the wall of the steel drums.

From the Editors

From the Editors

Preliminary Investigation on Atmospheric-Induced Stress Corrosion Cracking of British Intermediate Level Nuclear Waste Containers: Residual Stress Analysis and Four Point Bend Exposure Test

There are approximately 45,000 Intermediate Level Nuclear Waste (ILW) containers currently stored in the UK at Sellafield and elsewhere. These stainless steels containers should last up to 150 years of surface storage. In this study the general structural integrity of 500L ILW containers were investigated. The associated degradation issue under ambient conditions known as atmospheric-induced stress corrosion cracking (AISCC) has been studied. Preliminary investigation included profiling of residual stress distribution in parent and weld specimens, atmospheric exposure testing under various salts. So far residual stress profile shows high degree of complexity. No cracks have been observed until this moment in time.

Vacancy–solute interactions in ferromagnetic and paramagnetic bcc iron: Ab initio calculations

Vacancy–solute interactions play a crucial role in diffusion-controlled processes, such as ordering or decomposition, which occur in alloys under heat treatment or under irradiation. Detailed knowledge of these interactions is important for predicting long-term behavior of nuclear materials (such as reactor steels and nuclear-waste containers) as well as for advancing our general understanding of kinetic processes in alloys. Using first-principles calculations based on the density functional theory and employing the locally self-consistent Green’s function technique, we develop a database of vacancy–solute interactions in dilute alloys of bcc Fe with 3p (Al, Si, P, S), 3d (Sc–Cu), and 4d (Y–Ag) elements. Unrelaxed interactions within the first three coordination shells have been computed in the ferromagnetic state as well as in the paramagnetic (disordered local moment) state of the iron matrix. Magnetism is found to have a strong effect on the vacancy–solute interactions. Implications of the obtained results for interpreting the effects of vacancy trapping and enhanced impurity diffusion are discussed.

From Nanocomposite Aerogels to Glass Ceramics for Nuclear Wastes Containment

Nanocomposite aerogel is proposed as a host matrix for the synthesis of glass ceramics. The large porosity is used as a sponge to incorporate chemical species getting a two phases material. We describe the steps of the synthesis of glass ceramics for nuclear waste containment, from nanocomposite aerogels loaded with actinides surrogates (Ce and Nd). The glass synthesis is obtained without melting, by a control of several solid phase transformations: sintering, viscous flow, crystallization and foaming. Thanks to their high resistance to thermal shock and water corrosion, these glass ceramics are certainly good candidates as actinides containment materials.

Dependence of Vacancy-Solute Interactions on Magnetic State in Dilute Iron-Based Alloys

Vacancy-solute interactions play a crucial role in diffusion-controlled phase transformations, such as ordering or decomposition, which occur in alloys under heat treatment or under irradiation. The knowledge of these interactions is important for predicting long-term behavior of nuclear materials (such as reactor steels and nuclear-waste containers) under irradiation, as well as for advancing our general understanding of kinetic processes in alloys. Using first-principles calculations based on density functional theory and employing the locally self-consistent Green’s function technique, we develop a database of vacancy-solute interactions in dilute alloys of bcc Fe with 3p (Al, Si, P, S), 3d (Ti – Cu), and 4d (Nb – Ag) elements. Interactions within the first two coordination shells have been computed in the ferromagnetic state as well as in the paramagnetic (disordered local moment) state of the iron matrix. Magnetism is found to have a very strong effect on the vacancy-solute interactions.

Theoretical investigation of the evolution of the passive state on Alloy 22 in acidified, saturated brine under open circuit conditions

It is observed in many systems that as a metal or alloy is exposed to an aqueous solution over a long period of time, the corrosion potential drifts in the positive direction and the corrosion rate decreases. This implies the possibility that the potential may become sufficiently positive that it will exceed the transpassive potential and result in depassivation of the surface and hence to massive corrosion. On the other hand, if depassivation does not occur, the corrosion rate is predicted to continuously decrease with time, such that experiments carried out over short periods of time may seriously over-estimate the true corrosion rate over extended periods. This is an important issue in the selection of materials for the containment of high-level nuclear waste (HLNW), because of the excessively long performance horizon (up to 10 6 years), and in interpreting corrosion data from archeological artifacts. The theory developed in this work provides a physico-electrochemical explanation of the ennoblement phenomenon that is in good agreement with experiment and, in principle, allows one to optimize the model on short-term data and then predict the corrosion behavior over the very long term.

Effect of temperature on corrosion parameters and apparent activation energy measured by embedded sensors in pilot container in El Cabril repository

In order to study the real on-site aging of concrete and the potential corrosion risk for the reinforcements, Enresa has undertaken a monitoring programme in collaboration with Institute of Construction Science ‘Eduardo Torroja’ (IETcc) by permanently monitoring a pilot nuclear waste container through the installation of sensors. The pilot container of medium and low radioactive waste is buried near the real large concrete cells and has been monitored from 1994. The evolution from 1995 of corrosion parameters (corrosion rate, corrosion potential, electrical resistivity, concrete strains, oxygen availability) obtained is commented in the paper. The results indicate that the parameters evolve with time and the temperature is a very relevant variable influencing the measurements. An attempt to model the effect of temperature has been made by calculating the ‘apparent activation energies’ for the case of the electrical resistivity. Two clear regimes or slopes were found, one below and the other above 20°C, which have been attributed to a change in the state of the water from liquid to vapour although it has not been confirmed. The fact is that the resistivity increase is more pronounced above 20°C than below.

Nuclear waste management using alpha particle physical phenomena by nanoscale investigations

Nuclear waste is investigated from the aspect of its nanoscale behaviour. Four materials are selected as the nuclear waste container. Using the irradiation-induced amorphisation, some characteristics are examined. The Displacement Per Atom (dpa) is affected by the ion dose using the Stopping and Range of Ions in Matter 2008 (SRIM 2008) code system, which is a computer package of molecular dynamic simulations. The dpa is changed completely and kinetic energy is transferred to the target by the nuclear collision. The length of the material is a function of the ion collisions. It is concluded that a thickness of 204 nm is the optimised length of a waste drum by crystalline silicotitanate.

Safety implication for an unsaturated zone nuclear waste repository

In this paper we shed some light on the safety of unsaturated zone nuclear geological repositories in the long run by examining the effect of physical and chemical processes that take place inside a partially failed nuclear waste container. Our analysis addresses the safety of the proposed nuclear repository at Yucca Mountain, which is intended to store high-level nuclear waste. Our study is independent of the US Department of Energy (DOE) analysis, which involves a number of complex computer codes and assumptions, and relies on the performance of an engineered barrier system. Our safety analysis could be applied in general to any geological repository designed to be in an unsaturated zone, since it is based on the geology, unsaturated zone location, and a key characteristic of the waste, heat production. This analysis shows that the radionuclide release from a partially failed waste container, stored in an unsaturated zone geological repository, is likely to be gradual and long delayed.

Structural modifications induced by swift heavy ions in thin films of yttria fully stabilized zirconia

Among ceramic materials for nuclear waste containment, single crystal yttria fully stabilized zirconia (FSZ) gained particular consideration because of its excellent radiation resistance both in the elastic and inelastic collision regime. We deposited amorphous and polycrystalline, cubic FSZ thin films on (1 0 0) Si by ultraviolet pulsed laser ablation and irradiated them with swift heavy uranium ions of 2.6-GeV energy at fluences between 2 and 12 × 10 11 ions cm −2. The films were characterized before and after irradiation using X-ray reflectivity, grazing incidence X-ray diffraction, micro-Raman spectroscopy and transmission electron microscopy. Under ion irradiation, as-deposited crystalline films undergo amorphisation, followed by partial recrystallisation, whereas as-deposited amorphous films retain their disordered character. The dominant defects produced in the films are oxygen vacancies which may explain the amorphisation to recrystallisation path of our crystalline films.

Structural identification of a trioctahedral smectite formed by the aqueous alteration of a nuclear glass

The precipitation of crystallized magnesium phyllosilicates appears to control the residual alteration rate, under aqueous oxidizing conditions, of nuclear waste containment glass from the French Vitrification Facility at Marcoule (AVM). As a consequence, the precise identification of the formed phyllosilicates is crucial for modeling the long-term behavior of glass. Morphological and chemical analysis (scanning electron microscopy, EDS, ICP-AES after caustic fusion), and structural analysis (X-ray diffraction, MAS NMR) were used to determine the nature and structure of the precipitated phase: a trioctahedral smectite intermediate between saponite and hectorite, with the formula: Na 0.39(Mg 2.25Li 0.06Al 0.38Fe 0.06 M 0.25)(Si 3.23Al 0.77)O 10(OH) 2.

Determination of Crevice Corrosion Susceptibility of Alloy 22 Using Different Electrochemical Techniques

AbstractAlloy 22 belongs to the Ni-Cr-Mo family and it is highly resistant to general and localized corrosion. It may suffer crevice corrosion in aggressive environmental conditions. This alloy has been considered as a corrosion-resistant barrier for high-level nuclear waste containers. It is assumed that localized corrosion may occurs when the corrosion potential (ECORR) is equal or higher than the crevice corrosion repassivation potential (ER,CREV). The latter is measured by means of different electrochemical techniques using artificially creviced specimens. These techniques include cyclic potentiodynamic polarization (CPP) curves, Tsujikawa-Hisamatsu electrochemical (THE) method or other non-standard methods, such as the PD-GS-PD technique.The aim of the present work was to determine reliable critical or protection potentials for crevice corrosion of Alloy 22 in pure chloride solutions at 90°C. Conservative methodologies (which include extended potentiostatic steps) were applied for determining protection potentials below which crevice corrosion cannot initiate and propagate. Results from PD-GS-PD technique were compared with those from these methodologies in order to assess their reliability. Results from the CPP and the THE methods were also considered for comparison. The repassivation potential resulting from the PD-GS-PD technique was conservative and reproducible, and it did not depend on the amount of previous crevice corrosion propagation.

Iron Oxidation at Low Temperature (260–500 °C) in Air and the Effect of Water Vapor

The oxidation of iron has been studied at low temperatures (between 260 and 500 °C) in dry air or air with 2 vol% H2O, in the framework of research on dry corrosion of nuclear waste containers during long-term interim storage. Pure iron is regarded as a model material for low-alloyed steel. Oxidation tests were performed in a thermobalance (up to 250 h) or in a laboratory furnace (up to 1000 h). The oxide scales formed were characterized using SEM-EDX, TEM, XRD, SIMS and EBSD techniques. The parabolic rate constants deduced from microbalance experiments were found to be in good agreement with the few existing values of the literature. The presence of water vapor in air was found to strongly influence the transitory stages of the kinetics. The entire structure of the oxide scale was composed of an internal duplex magnetite scale made of columnar grains and an external hematite scale made of equiaxed grains. 18O tracer experiments performed at 400 °C allowed to propose a growth mechanism of the scale.

Water Corrodes Copper

According to a current concept, copper canisters of thickness 0.05 m will be safe for nuclear waste containment for 100,000 years. We show that more than 1 m copper thickness might be required for 100,000 years durability based on water exposures of copper for 20 h, 7 weeks, 15 years, and 333 years. An observed evolution of hydrogen which involves heterogeneous catalysis of molecular hydrogen, first principles simulations, thermodynamic considerations and corrosion product characterization provide further evidence that water corrodes copper resulting in the formation of a copper hydroxide. These findings cast additional doubt on copper for nuclear waste containment and other important applications.

Microbiologically Influenced Corrosion of Nuclear Waste Containers

Abstract Microbiologically influenced corrosion (MIC) is one of a number of threats to the long-term integrity of nuclear waste containers. As such, the potential for, and extent of, MIC must be assessed and suitable models developed for predicting the long-term behavior of the container. There are two broad approaches to assessing the threat posed by MIC; first, to determine whether the environment will support microbial activity and, if so, where and when it will occur, and second, to estimate the maximum amount of damage that could occur if microbial activity in the repository is possible. A decision-tree approach is used to present evidence for both of these approaches and to decide whether MIC is a significant threat to the integrity of the container. Examples are provided from various international nuclear waste management programs. It is concluded that microbial effects will not compromise the safety of the overall disposal system because they will not lead to either early container failures or to ...

Redox behavior of platinum-group metals in nuclear glass

Abstract This study highlights the role of two platinum-group metals, ruthenium and palladium, in the redox equilibria of nuclear waste containment glass. Electrochemical measurements in simplified R7/T7 glass melts were used to develop a thermodynamic model of ruthenium redox equilibrium. The oxygen fugacity at equilibrium, corresponding to the coexistence of ruthenium oxide and ruthenium metal dispersed in the molten glass, was measured at different ratios at temperatures ranging from 1000 °C to 1200 °C. Experiments were carried out on glass with and without Pd, revealing the combined role of palladium and tellurium on redox equilibria in the glass. The formation of palladium–tellurium alloys in nuclear glass was observed to result in oxidation of the elements dissolved in the melt.

Modifications of yttria fully stabilized zirconia thin films by ion irradiation in the inelastic collision regime

Yttria fully stabilized zirconia (FSZ) is a candidate material for nuclear inert matrix fuel cell and nuclear waste containment due to its isostructure with UO2 and PuO2 and its outstanding radiation resistance. Amorphous and polycrystalline cubic FSZ thin films of thickness around 400 nm were deposited on (100) Si by ultraviolet pulsed laser ablation and irradiated with 2.6 GeV uranium ions at fluences between 2×1011 and 1.2×1012 ions cm−2. The films were characterized before and after irradiation using scanning electron microscopy, atomic force microscopy, grazing incidence x-ray diffraction, and x-ray photoelectron spectroscopy (XPS). Amorphization, followed by partial recrystallization, is observed for irradiated crystalline films, whereas the amorphous films remain unaltered. A shift in the relative position of the XPS Zr 3d, Y 3d, and O 1s core lines is observed upon irradiation both in the crystalline and amorphous films, indicating differences in the local chemical environment at the surface as well as in near-surface layers. Such changes are ascribed to oxygen migration at the film surface, which may promote the recrystallization of as-deposited crystalline films but does not affect amorphous films.

Th diffusion in monazite

Diffusion of Th has been characterized in synthetic monazite under dry conditions. The synthetic monazites (either pure CePO 4, NdPO 4, or a mixed LREE phosphate containing Ce, Nd, and Sm) were grown via a Na 2CO 3–MoO 3 flux method. The sources of diffusant for the experiments were either synthesized ThSiO 4 or CaTh(PO 4) 2 powders mixed with CePO 4. Experiments were performed by placing source and monazite in Pt capsules and annealing capsules in 1-atm furnaces for times ranging from 3 weeks to a few hours, at temperatures from 1350 to 1550 °C. The Th distributions in the monazite were profiled by Rutherford Backscattering Spectrometry (RBS). The following Arrhenius relation was obtained for Th diffusion in monazite: D = 6.7 × 10 1 exp ( − 740 ± 63 kJ mol − 1 / R T ) m 2 s − 1 The diffusivity of Th was found to be similar for monazites containing a single REE (Ce or Nd) and the mixed LREE phosphates. Th diffusion was also similar for experiments run using the Th silicate and Ca–Th phosphate sources, suggesting that the substitutional mechanism for Th in monazite, i.e., Th + 4 + Si + 4 → REE + 3 + P + 5 for the ThSiO 4 source, and Th + 4 + Ca + 2 → 2REE + 3 for the CaTh(PO 4) 2 source, does not significantly affect Th diffusivities, and that Th is likely the rate-limiting species in diffusion. Th diffusion in monazite is about 3 orders of magnitude slower than Pb diffusion [Cherniak, D.J., Watson, E.B., Grove, M., Harrison, T.M., 2004a. Pb diffusion in monazite: a combined RBS/SIMS study. Geochimica et Cosmochimica Acta 68, 829–840.]. This contrasts with findings of Gardés et al. [Gardés, E., Jaoul, O., Montel, J.-M., Seydoux-Guillaume, A.-M., Wirth, R., 2006. Pb diffusion in monazite: An experimental study of Pb 2 + + Th 4+ → 2Nd 3+ interdiffusion. Geochimica et Cosmochimica Acta 70, 2325–2336.] who determined that diffusivities measured in (Pb, Th) ↔ REE exchange (which would under most circumstances be rate-limited by the slowest-diffusing species) are comparable to those measured by Cherniak et al. [Cherniak, D.J., Watson, E.B., Grove, M., Harrison, T.M., 2004a. Pb diffusion in monazite: a combined RBS/SIMS study. Geochimica et Cosmochimica Acta 68, 829–840.] for Pb. Th diffusion in zircon [Cherniak, D.J., Hanchar, J.M., Watson, E.B., 1997. Diffusion of tetravalent cations in zircon. Contributions to Mineralogy and Petrology 127, 383–390.] is about an order of magnitude slower than in monazite, but with similar activation energy for diffusion. The lower diffusivities in zircon may be a consequence of the larger disparity in size between Th and the Zr site in zircon as compared with Th and the REE site in monazite. Nonetheless, Th is essentially immobile in monazite with respect to exchange by volume diffusion under most geologic conditions. These findings may also have implications for containment of high-level actinide-based nuclear waste in monazite ceramic waste forms, but it should be noted that mechanisms other than diffusive exchange (e.g., recrystallization, dissolution/reprecipitation) may be important to consider along with diffusion when evaluating the retentivity of monazite for Th and other actinide elements.