High-level Waste Containers Research Articles

Modelling Hydraulic and Swelling Pressure Characteristics of Bentonite Buffer Material for High-Level Nuclear Waste Containment Conditions

Modelling Hydraulic and Swelling Pressure Characteristics of Bentonite Buffer Material for High-Level Nuclear Waste Containment Conditions

Perspective on “An Electrochemical Approach to Predicting Long-Term Localized Corrosion of Corrosion-Resistant High-Level Waste Container Materials,” D.S. Dunn, G.A. Cragnolino, and N. Sridhar, Corrosion 56, 1 (2000): p. 90-104

Perspective on “An Electrochemical Approach to Predicting Long-Term Localized Corrosion of Corrosion-Resistant High-Level Waste Container Materials,” D.S. Dunn, G.A. Cragnolino, and N. Sridhar, <i>Corrosion</i> 56, 1 (2000): p. 90-104

Operando scanning electron microscopy platform for in situ imaging of fluid evolution in nanoporous shale.

Fluid-solid interactions in nanoporous materials underlie processes fundamental to natural and engineered processes, including the thermochemical transformation of argillaceous materials during high-level nuclear waste disposal. Operando fluid-solid resolution at the nanoscale, however, is still not possible with existing optical and electron microscopy approaches that are constrained by the diffraction limit of light and by vacuum-fluid incompatibility, respectively. In this work, we develop an operando scanning electron microscopy (SEM) platform that enables the first direct in situ imaging of dynamic fluid-solid interactions in nanoporous materials with spatio-temporal-chemical resolutions of ∼2.5 nm per pixel and 10 fps, along with elemental distributions. Using this platform, we reveal necessary conditions for thermochemical pore and fracture generation in shales and measure their surface wetting characteristics that constrain the feasibility of high-level nuclear waste containment. Notably, we show that low heating-rate conditions typical of radioactive decay produce hydrocarbon liquids that wet fracture and pore surfaces in a self-sealing manner to impede aqueous radionuclide advection.

Corrosion behavior of high-level waste container materials Ti and Ti–Pd alloy under long-term gamma irradiation in Beishan groundwater* *Project supported by the National Natural Science Foundation of China (Grant Nos. 51471160, 11775102, and 11965001) and the Fundamental Research Funds for the Central Universities, China (Lanzhou University, Grant No. lzujbky-2018-19).

Titanium and titanium–palladium alloys are important potential materials for nuclear waste container, which will endure both intense γ-irradiation and groundwater erosion. Therefore, it is very important to investigate the corrosion behavior of the container materials. In this research, the cumulative dose effect of TA8-1 type titanium–palladium alloy (TA8-1) and TA2-type pure titanium (TA2) under γ-irradiation was studied based on the geological disposal of nuclear wastes. The irradiation experiments were performed at room temperature using 60Co gamma sources with a 5.0-kGy⋅h−1 intensity for 40, 80 or 160 days, respectively. The pH value and conductivity of Beishan groundwater were investigated. The results showed that the pH value changed from alkaline (8.22) to acidic (2.46 for TA8-1 and 2.44 for TA2), while the un-irradiated solution remained alkaline (8.17 for TA8-1 and 8.20 for TA2) after 160 days. With the increase of irradiation dose, the conductivity increases rapidly and then tends to become stable, which indicates that the titanium dioxide corrosion layer formed on the surface of the sample surface effectively prevents further corrosion. Meanwhile, XRD and SEM–EDS analysis results show that the main components of corrosion products are TiO2 and TiO. The titanium on the surface of the sample is oxidized, resulting in slight uneven local corrosion. The results show that TA8-1 and TA2 are suitable to be used as candidate materials for high-level waste (HLW) disposal containers due to their excellent performance under long-term and high-dose irradiation corrosion.

Effect of cooling profile on crystalline phases, oxidation state, and chemical partitioning of complex glasses

Abstract:Investigations of the crystallization of aluminosilicate phases within Hanford nuclear waste glasses typically involve subjecting samples to the canister centerline cooling (CCC) schedule. This cooling schedule is representative of the slowest cooling thermal profile which these glasses will experience after the glass is poured into the high level waste (HLW) container. However, few investigations have observed how the crystallization behavior changes by varying the heat treatment schedule. In the present study, three Hanford HLW glasses are subjected to CCC and isothermal heat treatments (IHT) to better understand the evolution of phases and the chemical partitioning due to temperature schedule. Samples were characterized using electron probe microanalysis, X-ray diffraction, micro X-ray fluorescence, and micro X-ray absorption spectroscopy. From IHT, eucryptite and apatite phases were observed which were not observed during CCC. Spatially-resolved measurements demonstrated that the oxidation state of the iron was similar among glass and crystal, and we suggest a mechanism to describe the compositional fluctuations near the crystal-glass interface which influence crystallization.

Radionuclide adsorption mechanism in buffer materials in high-level radioactive waste container: MD study [高放废物深地质处置库蒙脱土对铀酰的吸附阻滞行为: 分子模拟研究

Radionuclide adsorption mechanism in buffer materials in high-level radioactive waste container: MD study [高放废物深地质处置库蒙脱土对铀酰的吸附阻滞行为: 分子模拟研究

Conceptual design of ultra-high performance fiber reinforced concrete nuclear waste container

Abstract This research presents a structural design of high-level waste (HLW) container using ultra-high performance fiber reinforced concrete (UHP-FRC) material. The proposed design aims to overcome the drawbacks of the existing concrete containers which are heavy, difficult to fabricate, and expensive. In this study, the dry storage container (DSC) that commonly used at Canadian Nuclear facilities is selected to present the proposed design. The design has been performed such that the new UHP-FRC alternative has a structural stiffness equivalent to the existing steel-concrete-steel container under various loading scenarios. Size optimization technique is used with the aim of maximizing stiffness, and minimizing the cost while satisfying both the design stresses and construction requirements. Then, the integrity of the new design has been evaluated against accidental drop-impact events based on realistic drop scenarios. The optimization results showed: the stiffness of the UHP-FRC container (300 mm wall thick) is being in the range of 1.35–1.75 times the stiffness of existing DSC (550 mm wall thick). The use of UHP-FRC leads to decrease the container weight by more than 60%. The UHP-FRC container showed a significant enhancement in performance in comparison to the existing DSC design under considered accidental drop impact scenarios.

Characterisation of Excavation-Induced Damage Around a Short Test Tunnel in the Opalinus Clay

In Switzerland, the Opalinus Clay is under consideration as a potential host rock for deep geological storage of nuclear waste. The Swiss concept involves high-level waste containers emplaced in small-diameter drifts of roughly 3 m. At the Mont Terri Underground Rock Laboratory, a short mine-by experiment (EZ-B) was executed in 2005 with the objective of characterising the excavation-induced damage through an interpretation of integrated field data. The damage zone was found to consist of two parts. Fracture mapping from drillcores and televiewer images revealed a thin inner zone of macroscopic fracturing (20 cm) and devoid of borehole instabilities as well as coinciding with the lowest P-wave characteristics (normalised amplitudes and velocities). The outer zone (50–80 cm) was defined by increasing P-wave characteristics and evidence of borehole instabilities. Supplementing the data interpretation, a simplified numerical elastic stress analysis indicated that the rock mass in the sidewalls and upper western haunch are the most susceptible to spalling. In these regions, stress levels only just reach a spalling limit around 0.05 in the inner fractured zone, suggesting that the zone of macro-fracturing around the niche is not significant. It is postulated that the physical manifestation of the relatively weak bedding plane strength is dominated by bedding-perpendicular displacement as opposed to bedding-parallel shear.

Corrosion of single layer thin film protective coatings on steel substrates for high level waste containers

Single-layer thin film coatings have been deposited on steel substrates and tested for their corrosion resistance. These coatings include TiN, ZrO2, TiO2, Al2O3, and MoS2, and it is proposed that they will act as barriers to provide protection to the steel canisters that are part of the dry cask storage system for high level nuclear waste. Corrosion testing was completed using electrochemical potentiodynamic polarization techniques in aerated 1 M NaCl solution. Results show an exponential increase in corrosion rate with increasing temperature and an exponential decrease in the passive breakdown overpotential, which is directly related to the ability of a material to form and sustain a corrosion-inhibiting passive film in a given environment. Additionally, kinetic activation parameters have been experimentally determined for each material, leading to predictive equations for corrosion rates. The bare and coated samples corrode analogously, indicative of pores allowing the coating and substrate to corrode simultaneously. The samples were also placed in circulating salt brines of varying pH as a supplementary corrosion testing mechanism to explore their corrosivity over extended time. Negligible weight change was experienced by the bare and coated steel samples over a period of 5 months. Increasing the coating thickness and the number of layers may provide higher resistance to uniform and localized corrosion.

Shielding properties of protective thin film coatings and blended concrete compositions for high level waste storage packages

Various thin film coatings have been proposed to protect stainless steel high level waste (HLW) containers from premature failure due to localized corrosion, hydrogen embrittlement, and mechanical wear. These coatings include TiN, ZrO2, MoS2, TiO2, and Al2O3, to be deposited either in multiple layers or as a thicker, single-layer composite. Linear attenuation coefficients of these materials have been simulated using MicroShield and measured experimentally for various photon energies. Additionally, spent fuel casks with overpacks made of two different types of concrete were simulated to compare exposure rate at the cask surface. In the energy range that is significant for high level waste storage all coating materials possess very similar attenuation behavior. A specialty concrete, containing magnetite (Fe3O4) and lead oxide (PbO), reduces the exposure rate at the outer surface of the overpack by several orders of magnitude. The higher-Z elements not present in ordinary concrete greatly increase attenuation of intermediate-energy gammas (0.4–1.0MeV). The thin film coatings do not affect the shielding capabilities of the HLW packaging, as their total proposed thickness is nearly three orders of magnitude less than the mean free path (MFP) of the primary photons of interest.

The ISSMGE Kerry Rowe Lecture: The role of diffusion in environmental geotechnics

Diffusion of contaminants can play a significant if not dominant role in many applications encountered within the field of environmental geotechnics. The objective of this paper is to provide an overview of the important role diffusion plays in such applications. The presentation proceeds from a historical perspective, beginning with the recognition in the late 1970s to early 1980s that diffusion may be an important process in assessing contaminant migration through low-permeability barriers in waste containment applications. Data from the literature and simplified model simulations are used to illustrate under what conditions diffusion is important, and the significance of diffusion is illustrated with respect to different barrier components and types of barriers used in waste containment applications. The barriers considered include natural clays, compacted clay liners, geomembranes, geosynthetic clay liners, composite liners, vertical cutoff walls, subaqueous caps for contaminated sediments, and highly compacted bentonite buffers for high-level radioactive waste containment. The significance of semi-permeable membrane behavior on liquid-phase diffusion through bentonite-based barriers also is highlighted. The potential importance of matrix diffusion as an attenuation mechanism for contaminant transport is illustrated, and the roles of both liquid-phase and gas-phase diffusion under unsaturated conditions are discussed. Finally, the role of diffusion in terms of remediation applications is illustrated via an example analysis illustrating the impact of reverse matrix or back diffusion on the effectiveness of pump-and-treat remediation, as well as via a summary of several diffusion-based models commonly used to describe the leaching of contaminants from a variety of stabilized–solidified waste forms.

Investigation of corrosion behaviours of high level waste container materials in simulated groundwater in China

Appropriate high level radioactive waste container materials in China are still under investigation. The corrosion behaviours of four types of materials (carbon steel Q235, aluminium alloy 2024, copper and stainless steel SUS 304) in simulated groundwater at room temperature were investigated. The corrosion rate was measured through the weight loss method. The chemical composition of corrosion product films formed on the surface of materials was investigated by means of X-ray diffraction and energy dispersive X-ray spectroscopy. The results showed that there was no crevice and pitting corrosion for carbon steel after one year of immersion; the corrosion was dominated by general corrosion, the composition of the corrosion product film being mainly γ-FeOOH, as well as Fe3O4. Copper underwent slight crevice corrosion, but aluminium alloy 2024 suffered serious localised corrosion. No corrosion was observed on stainless steel SUS 304. The presence of oxygen promoted crevice corrosion of copper and aluminium alloy 2024.

Mechanism of RuO2Crystallization in Borosilicate Glass: An Originalin SituESEM Approach

Ruthenium, a fission product arising from the reprocessing of spent uranium oxide (UOX) fuel, crystallizes in the form of acicular RuO(2) particles in high-level waste containment glass matrices. These particles are responsible for significant modifications in the physicochemical behavior of the glass in the liquid state, and their formation mechanisms are a subject of investigation. The chemical reactions responsible for the crystallization of RuO(2) particles with acicular or polyhedral shape in simplified radioactive waste containment glass are described. In situ high-temperature environmental scanning electron microscopy (ESEM) is used to follow changes in morphology and composition of the ruthenium compounds formed by reactions at high temperature between a simplified RuO(2)-NaNO(3) precursor and a sodium borosilicate glass (SiO(2)-B(2)O(3)-Na(2)O). The key parameter in the formation of acicular or polyhedral RuO(2) crystals is the chemistry of the ruthenium compound under oxidized conditions (Ru(IV), Ru(V)). The precipitation of needle-shaped RuO(2) crystals in the melt might be associated with the formation of an intermediate Ru compound (Na(3)Ru(V)O(4)) before dissolution in the melt, allowing Ru concentration gradients. The formation of polyhedral crystals is the result of the direct incorporation of RuO(2) crystals in the melt followed by an Ostwald ripening mechanism.

Modeling and Measurement of Residual Stresses in a Steel Vessel Containing Glass

Residual stresses in a stainless steel vessel containing glass have been evaluated using measurements and numerical simulation. High-level nuclear wastes are often vitrified in glass cast in cylindrical stainless steel containers. Knowledge of the internal stresses generated in both the glass and container during this process is critical to structural integrity and public safety. In this research, residual stresses were measured near the surface of a High Level Waste container using an Incremental Center Hole Drilling technique. Residual stress magnitudes were found to be at or near to the yield stress in the container wall. A transient finite-element thermal-stress model has been developed to simulate temperature, distortion, and stress during casting and cooling in a simple slice domain of both the glass and the container. Contact thermal-stress elements were employed to prevent penetration at the glass–container interface. Roughness of these contact surfaces was modeled as an equivalent air gap with temperature-dependent conductivity in the thermal model. The stress model features elastic-viscoplastic constitutive equations developed based on the temperature-dependent viscosity of the glass and elastic-plastic constitutive equations for the stainless steel. The simulation was performed using the commercial ABAQUS program with a user material subroutine. The model predictions are consistent with the residual stress measurements, and the complete thermal–mechanical behavior of the system is evaluated.

Metallic corrosion processes reactivation sustained by iron-reducing bacteria: Implication on long-term stability of protective layers

In deep geological environments foreseen for the disposal of radioactive waste, metallic containers will undergo anaerobic corrosion. In this context, the formation of corrosion products such as magnetite may reduce the rate of corrosion processes through the formation of a protective layer. This study aims at determining the direct impact of iron-reducing bacteria (IRB) activity on the stability of corrosion protective layers. Batch experiments investigating iron corrosion processes including the formation of secondary magnetite and its subsequent alteration in the presence of IRB show the bacteria ability to use structural Fe(III) for respiration which leads to the sustainment of a high corrosion rate. With the bio-reduction of corrosion products such as magnetite, and H 2 as electron donor, IRB promote the reactivation of corrosion processes in corrosive environments by altering the protective layer. This phenomenon could have a major impact on the long-term stability of metallic compounds involved in multi-barrier system for high-level radioactive waste containment.

Kinetics of AgI precipitation from AgCl as affected by background electrolyte

Iodide retention by AgCl, a potential sorbent in high-level waste (HLW) storage systems, was determined. The kinetics and steady state sorption of iodide were determined in single and mixed electrolytes of NaNO3, and NaCl at ionic strengths of 25 and 50 mM. Iodide retention involved the conversion of AgCl to AgI. This conversion increased rapidly within 0.02 hours, and retention maxima of 0.92 and 1.0 mol·l·mol−1 Ag occurred for low and high ionic strengths, respectively. These short-term studies indicated that AgCl would be an effective scavenger of I− in HLW containment systems.

Long-Term Integrity of Waste Package Final Closure for HLW Geological Disposal, (IV) Influence of Welding and Prediction of Long-Term Integrity of Weld Joint

Overpack, a high-level radioactive waste container for geological disposal in the Japanese waste management program, is required to isolate the sealed vitrified waste from contact with groundwater for 1,000 years. This means that the weld joint between the body and the lid of the overpack is also required to endure for this period of time. The authors have already started studies on a methodology guaranteeing the long-term integrity of the lid closure, the applicability of several welding and non-destructive examination methods for the weld joint and the corrosion properties of the weld joint. In this paper, through review of the obtained test results, the influences of welding were summarized. Considering the presence of weld flaws, crack growth phenomenon was estimated by fracture mechanics assessment, while the critical crack length and the maximum tolerable flaw size were discussed. Moreover, referring to the Fitness for Service for nuclear power plant components, a method guaranteeing the long-term integrity of the lid closure was discussed and proposed with consideration to the physical and chemical environmental conditions peculiar to the overpack.

Irradiation stability of R7T7-type borosilicate glass

High-level nuclear waste containment glass is subjected to irradiation-induced stresses whose consequences must be assessed to guarantee the material behavior over time. Alpha decay from the minor actinides confined in the glass structure is responsible for most of the atom displacements. Minor actinide concentrations in the glass exceeding current levels are under consideration for future vitrification scenarios. Four R7T7-type borosilicate glass samples doped with 0.04, 0.4, 1.2 and 3.25 wt% 244CmO 2 were fabricated to study not only the cumulative effect of high alpha decay doses (>10 19 α/g) but also of the dose rate. The homogeneity of the materials and their chemical compositions were characterized to ensure they were suitably representative for the experimental program. The macroscopic behavior of the glasses was characterized for doses up to 3 × 10 18 α/g. No significant effect on the initial alteration rate was detected which means that the chemical reactivity of the glass with pure water is not affected by such levels of alpha doses. The glass swelled slightly depending on the dose, saturating at about 0.5% after receiving a dose of about 2 × 10 18 alpha disintegrations per gram of glass. The mechanical properties were observed to improve appreciably, with lower hardness but greater fracture toughness. A stabilization phenomenon comparable to that of the glass density was also observed. Comparing the experimental data with other findings obtained by atomistic modeling and external irradiation suggests that nuclear interactions caused by alpha decay recoil nuclei could be responsible for the experimental variations observed. The possible origins of the observed variations in macroscopic properties are also discussed.

A Review of 25 years of Corrosion Studies on HLW Container Materials at the CEA

ABSTRACTThe Commissariat à l'Énergie Atomique (CEA, French Atomic Energy Commission) has been involved in researches on nuclear waste management for more than 25 years. One of the key issues is the prediction of the long term behavior and aging of the High Level Waste (HLW) containers in order to develop concepts that will ensure the confinement of the activity over extremely long periods of time.Preliminary studies were carried out on two concepts, one of a thin "corrosion resistant" container made of titanium or nickel base alloy, and the other on a thick "corrosion allowance" container made or carbon steel.The results of these experiments showed that the "corrosion resistant" concept led to a high uncertainty on the development and propagation rate of localized forms of corrosion, and the concept of geological disposal in an argillaceous host formation of thick waste containers made of carbon steel was chosen as the reference. Studies are still in progress to better assess the corrosion mechanisms relevant to this situation in order to provide reliable models for the long term prediction of the containers corrosion behavior.

Influence of Welding and Heat Treatment on Corrosion of a High-Level Waste Container Material Carbon Steel in Disposal Salt Brines

Abstract Previous corrosion studies identified carbon steels as promising materials for the manufacture of long-lived, high-level waste containers that could act as an engineered barrier in a rock-salt repository. In the present work, the effect of electron beam (EB) and tungsten-inert-gas (TIG) welding (potential container closure techniques) on the corrosion behavior of the preselected, fine-grained structural steel TStE355 (DIN 1.0566) was investigated. The steel was examined in three material conditions such as unwelded, as-welded, and welded plus stress-relief thermal-treated in a disposal-relevant magnesium chloride (MgCl2)-rich brine at 150°C and a gamma radiation field of 10 Gy/h. Results indicated that the unwelded steel was resistant to pitting corrosion in the sense of an active-passive corrosion element, and its general corrosion implied corrosion allowances acceptable for thick-walled containers. The EB- and TIG-welding significantly decreased the corrosion resistance of the steel. The as-wel...