Carbon Steel Overpack Research Articles

The anaerobic corrosion of the carbon steel overpack under anoxic alkaline conditions representing the Belgian supercontainer concept

Abstract. The Belgian Agency for Radioactive Waste and Enriched Fissile Materials (ONDRAF/NIRAS) is responsible for the management of all radioactive materials (currently existing and future waste) generated in Belgium. Disposal in deep stable geological formations is, at present, accepted at an international level to be the most promising option for the long-term management of high-level and/or long-lived radioactive waste. ONDRAF/NIRAS currently recommends underground disposal in a geological stable and poorly indurated clay formation as the reference management option to steer their RD&D research programmes. The supercontainer concept is currently being considered as the reference design for the final disposal of vitrified high-level radioactive waste (VHLW) and spent fuel (SF) in Belgium. It comprises a prefabricated Portland cement-based buffer that completely surrounds a carbon steel overpack. In this highly alkaline environment (pH ∼ 13.6), and under normal conditions (i.e. without the ingress of aggressive species), the carbon steel overpack will be protected by a passive oxide film, which is expected to result in very low uniform corrosion rates. The main goal of the ongoing R&D corrosion studies performed at ONDRAF/NIRAS is to provide confidence that the integrity of the carbon steel overpack will not be jeopardized at least for the duration of the thermal phase. Considering the long timescales applicable to waste disposal, determining accurate and reliable estimates of the uniform corrosion rate under anoxic conditions forms a very important part of the R&D methodology of ONDRAF/NIRAS' corrosion programme. On the other hand, the occurrence of localized forms of corrosion cannot be neglected, and therefore it is also crucial to demonstrate that no other form of corrosion apart from uniform corrosion can take place. This paper gives an overview of the status of the research performed at ONDRAF/NIRAS with respect to the uniform corrosion, pitting corrosion and stress corrosion cracking behaviour of the carbon steel overpack of the supercontainer.

Experimental factors and their impact on understanding corrosion in anoxic environments

AbstractThe long‐term disposal of nuclear waste in deep geological repositories requires safe containment for up to one million years. To understand how the waste will respond to the gradually evolving environment, researchers can perform sensitive experiments in the laboratory to replicate the repository conditions at specific time points. In Belgium, spent nuclear fuel will be housed within a carbon steel overpack, which is encased within a cement buffer, which may include a stainless steel liner (the “envelope”). The range of corrosion behaviors for carbon steel, and to a lesser extent, stainless steel, has been studied through hydrogen evolution, an end‐product of oxidation under repository conditions. Considerable time may be required for metallic surfaces to approach a steady hydrogen evolution and this is dependent upon the starting condition of the specimen surfaces. The use of nominally‐similar materials can result in significant variation in the reported corrosion rates. The presence of cements, which generate their own hydrogen and influence water chemistry as a function of time, add complexity to even simple experimental configurations that requires careful interpretation.

Corrosion Behavior and Sacrificial Properties of Zn and Zn-Al Coatings in Conditions Simulating Deep Geological Disposal of Radioactive Waste at 80 °C

In France, it is planned to manage high-level radioactive waste via deep geological disposal. The carbon steel overpacks containing the waste will face corrosion processes specific to the unusual environment of the disposal. To protect these overpacks against corrosion, the use of metallic sacrificial coatings was considered as a possible solution. Therefore, the corrosion behavior of thermally sprayed Zn and Zn-15wt.%Al coatings (500 µm thick) was studied in a solution simulating the specific environment of the disposal. Galvanic corrosion experiments were performed with a coated steel electrode connected to a bare steel one, which simulated a coating defect. First, it was observed that the Zn coating had lost its sacrificial properties after ~2 months of coupling. XRD and µ-Raman analysis carried out after the experiment demonstrated that the coating was covered with a protective hemimorphite (Zn4Si2O7(OH)2·H2O)-rich layer. In contrast, the Zn-15wt.%Al coating kept its sacrificial properties all along the 6-month experiment. It proved to be covered with a layer mainly composed of Zn and Zn-Al hydroxysulfates, which did not contain hemimorphite. However, SEM cross-sectional observations showed that the electrolyte had seeped into the coating and had even reached the steel surface. This infiltration led to the oxidation of the splats inside the coating, even close to the steel surface, and induced a large swelling of the coating. Its thickness was increased by 50%.

Cement as a component of the multi-barrier system: radionuclide retention in cementitious environments

Abstract. Safety concepts regarding nuclear waste disposal in underground repositories generally rely on a combination of engineered and geological barriers, which minimize the potential release of radionuclides from the containment-providing rock zone or even their transport into the biosphere. Cementitious materials are used for conditioning of certain nuclear waste types, as components of waste containers and overpacks, as well as being constituents of structural materials at the interface between backfilling and host rock in some repository concepts. For instance, the preferred option for the disposal of high-level waste (HLW) in Belgium is based on the supercontainer design, which consists of a carbon steel overpack surrounded by a thick concrete buffer (Bel et al., 2006). In the event of formation water interacting with cementitious materials, pore water solutions characterized by (highly) alkaline pH conditions will form. These boundary conditions define the chemical response of the radionuclides, but also influence the behaviour of neighbouring components of the multi-barrier system, e.g. bentonitic or argillaceous backfilling and host rock. Hardened cement paste or Sorel cement are considered to be main sorbing materials present in the near field of repositories for low- and intermediate-level waste (L/ILW). Hence, interactions of radionuclides with cementitious materials represent a very important mechanism retarding their mobility and potential migration from the near field (Wieland, 2014; Ochs et al., 2016). While the quantitative description of the sorption processes (usually in terms of sorption coefficients, i.e. Kd values) is a key input in the safety analysis of nuclear waste repositories, detailed mechanistic analysis and understanding of sorption phenomena provide additional scientific arguments and important process understanding, and thus enhance both the quality of safety arguments and the overall confidence in the safety assessment process. Research at KIT-INE dedicated to the interaction of cementitious materials with radionuclides is conducted in the context of different repository concepts, including clay (low- and high-ionic strength conditions), crystalline rock or rock salt. Experimental and theoretical studies are performed within the framework of national (GRAZ, BMWi) and international (CEBAMA and EURAD-CORI, EU Horizon 2020 Programme) projects, extending to third-party projects with several waste management organizations in Europe, e.g. SKB (Sweden), ONDRAF-NIRAS (Belgium) or BGE (Germany). The combination of classical experimental (wet chemistry) methods, advanced spectroscopic techniques and theoretical calculations provides both an accurate quantitative evaluation and a fundamental understanding of the sorption processes. Examples of recent studies at KIT-INE on radionuclide behaviour in cementitious systems in the context of both L/ILW and HLW will be presented in this contribution to explain methodologies, scientific approaches and results. The present state of knowledge as well as main remaining uncertainties affecting the retention processes of radionuclides in cementitious environments under different conditions will be critically discussed, also in view of current international research activities and repository projects.

Further studies on the anaerobic corrosion of carbon steel in alkaline media in support of the Belgian supercontainer concept

AbstractIn the Belgian supercontainer concept, a carbon steel overpack surrounds high‐level radioactive waste and spent fuel containers and is encased in a cementitious buffer material. Research has been carried out to characterise the anaerobic corrosion of carbon steel in an anoxic artificial alkaline porewater that simulates the cementitious buffer material and also in solid cement matrices, in the presence of chloride and the sulphur species expected in the groundwater. The current study focuses on recent results. Specifically, it provides an update on the long‐term corrosion rates obtained from gas generation measurements (including new data from higher‐sensitivity autoclaves) and weight loss measurements and provides results of postmortem analysis of dismantled cells. The behaviour of the system is discussed, focusing on the effects of the solution chemistry, precorrosion treatments and γ radiation on corrosion rates, the electrochemical response of the steel when irradiated, and the morphology of the corrosion product. The corrosion rate was lower in cement than in an aqueous system. Radiation affected the corrosion potential and the morphology of the corrosion product.

Overview of anaerobic corrosion of carbon steel radioactive waste packages in alkaline media in support of the Belgian supercontainer concept

AbstractThe so‐called supercontainer concept is currently being considered as the reference design for the final disposal of vitrified high‐level radioactive waste and spent fuel in Belgium. It comprises a prefabricated Portland cement‐based buffer that completely surrounds a carbon steel overpack. In this highly alkaline environment (pH ~13.6), and under normal conditions (i.e., without the ingress of aggressive species), the carbon steel overpack will be protected by a passive oxide film, which is expected to result in very low uniform corrosion rates. Considering the long timescales applicable to waste disposal, determining accurate and reliable estimates of the uniform corrosion rate under anoxic conditions forms a very important part of the safety and feasibility strategy. It is also crucial to demonstrate that no other form of corrosion apart from uniform corrosion takes place. This paper presents the progress of the research that has been made since the LTC2016 Workshop (Toronto, Canada) on uniform corrosion, pitting corrosion and stress corrosion cracking behaviour of P355 QL2 grade carbon steel exposed to an artificial concrete pore water solution.

Electrochemical behavior of carbon steel with bentonite/sand in saline environment

AbstractCurrent designs for the geological disposal of high‐level radioactive waste in Japan use carbon steel overpack containers surrounded by a mixed bentonite/sand buffer material, which will be located in a purpose‐built repository deep underground. Though there are many suitable sites for a repository in Japan, the coastal areas are preferable from a logistical point of view. It is, therefore, important to evaluate the long‐term performance of the carbon steel overpack and the mixed bentonite/sand buffer material in the saline groundwaters of coastal areas. In the current study, the passivation behavior and initial corrosion rates of carbon steel with and without mixed bentonite/sand were tested as a function of the pH and salt concentration in representative saline groundwaters. The main findings indicate that the passivation of carbon steel encapsulated in a buffer material will be unlikely in a saline environment, even at high pH (12), and that the corrosion rate of carbon steel is more strongly affected by the presence of the buffer material than by the salt concentration.

Pitting of carbon steel in the synthetic concrete pore solution

AbstractPitting corrosion is a possible mode of failure of the carbon steel overpack of the Belgian supercontainer concept for the isolation of high‐level nuclear waste (HLNW). However, no firm experimental data are currently available to estimate the probability of failure over the extended storage time (100,000 years). Extensive work shows that passivity breakdown results from the condensation of cation vacancies (CVs) at the metal/barrier layer (m/bl) interface, in response to the absorption of Cl− into oxygen vacancies at the surface of the barrier oxide layer. The CVs migrate across the bl to the m/bl interface where they condense, leading to the separation of the bl from the metal. The resulting blister prevents the growth of bl into the metal and dissolution results in blister rupture, marking a passivity breakdown event. Stabilization via differential aeration produces a potentially damaging, stable pit. We review our work on passivity breakdown and the nucleation of pits on P355 QL2 carbon steel in high‐pH aqueous media typical of concrete pore solution, with emphasis on the mechanistic aspects. We conclude that failure of the carbon steel overpack containing the HLNW over a storage horizon of 100,000 years is improbable.

General corrosion of carbon steel in a synthetic concrete pore solution

AbstractIn this study, we reviewed our recent work on the general corrosion of carbon steel (P355QL2) overpack material for the isolation of high‐level nuclear waste in Belgium's supercontainer concept. By using electrochemical impedance spectroscopy and by optimizing the mixed potential model, which incorporates quantum mechanical tunneling of charge carriers across the barrier layer to describe the kinetics of the partial cathodic process, we evaluated all parameters in the model as a function of independent variables such as voltage, temperature, and pH. By delineating the partial anodic and cathodic processes, we found that the corrosion rate (CR) is independent of voltage over the voltage range from 0.2 to −1.0 VSHE, which is predicted to be experienced in the repository. Furthermore, the CR is found to increase strongly with decreasing pH and increasing temperature.

Monte-Carlo Simulation of Pitting Corrosion with a Deterministic Model for Repassivation

A theory is developed for calculating the delayed repassivation (“stifling”) rate constant for stable pits on a metal surface with attention being paid to pitting of the carbon steel overpack in the Belgian supercontainer concept of the disposal of high-level nuclear waste. It is shown that, under the conditions that are expected to exist at the overpack surface, substantial concentration and potential drops cannot exist between the pit internal surface and the external surface far away from the pit mouth. This conclusion is valid for both open and semi-closed pits, if there is no porous, high resistance, corrosion product layer (salt film and/or remnants of the ruptured passive film) between the active metal surface inside the pit and the external surface, as supported by recent artificial pit experiments. It is shown that the existence of a low conductivity, corrosion product layer between the active pit base and the external surface can protect pits from the repassivation. The condition of repassivation is based upon the concept of the existence a critical pit propagation rate, Vcr, corresponding to a critical coupling current. If the rate of pit propagation becomes less than Vcr the pit passivates, because the coupling current is too low to maintain the necessary aggressive conditions within the pit cavity. A model and computer code have been developed for estimating the probability of failure in the Belgian high-level nuclear waste (HLNW) disposal repository.

Influence of iron on the alteration of the SON68 nuclear glass in the Callovo-Oxfordian groundwater

In France, the high level radioactive waste, contained in a glassy matrix, is intended for a deep geological disposal. It is then important to understand glass alteration, which depends on the near-field materials and the environmental conditions. The corrosion of the carbon steel overpack (used in the French multi-barrier concept for the disposal of high level radioactive waste) could influence glass alteration by the release of iron in the solution and the formation of iron corrosion products. For a better understanding of these interactions, different experiments were performed at 90 °C using a SON68 glass (non-radioactive surrogate of the R7T7 glass) separated by 80 μm from a pristine iron foil and immersed in synthetic groundwater. A pre-corroded iron foil and a PTFE foil, replacing the pristine iron foil, were also used. The influence of iron on glass alteration was studied, focusing on two different environments (the confined environment, inside the crack, and the diluted environment, where the faces are exposed to the homogeneous solution) and on the influence of a pre-corrosion of the iron foil. Depending on the previously described conditions, some variations were observed: as a whole, glass alteration increased in the presence of iron due to the precipitation of Fe-silicate minerals and it was also higher in the diluted environment than in the confined environment, indicating that the gel protective properties were different in both cases. This might be related either to differences in the solution composition in contact with the glass or differences in the gel composition, as it was shown that iron can enter the porous gel layer. In the tested conditions, pre-corrosion of an iron foil seemed to have no influence on glass alteration in a confined environment.

Abiotically or microbially mediated transformations of magnetite by sulphide species: The unforeseen role of nitrate-reducing bacteria

In the anoxic conditions of nuclear waste deep geological repository, the main iron corrosion products formed at the surface of the carbon steel overpacks are siderite (FeIICO3) and magnetite (FeIIFeIII2O4) that form a rather protective layer. The originality of this work is the consideration of the bacterial metabolism of nitrate-reducing bacteria (NRB), often overlooked in corrosion processes. Klebsiella mobilis was used as a model of NRB and incubated with ferrous carbonates/magnetite corrosion bilayers under anoxic conditions. As a consequence of a combination of biotic and abiotic processes magnetite was reduced to mackinawite (FeIIS) and subsequently transformed to greigite (FeIIFeIII2S4).

Effect of Thiosulfate on the SCC Behavior of Carbon Steel Welds Exposed to Concrete Pore Water Under Anoxic Conditions

The Supercontainer (SC) is the reference concept for the post-conditioning of vitrified high-level nuclear waste and spent fuel in Belgium. It comprises a prefabricated concrete buffer that completely surrounds a carbon steel overpack. Welding is being considered as a final closure technique of the carbon steel overpack in order to ensure its water tightness. Welding is known to induce residual stresses near the weld zone, which may lead to an increased susceptibility to stress corrosion cracking (SCC). In this study, slow strain rate tests were conducted to study the SCC behavior of plain and welded P355 QL2 grade carbon steel exposed to an artificial concrete pore water solution that is representative of the SC concrete buffer environment. The tests were performed at 140°C, a constant strain rate of 5 × 10-7 s-1 and at open circuit potential under anoxic conditions. The effect of thiosulfate on the SCC behavior was investigated up to levels of 600 mg/L S2O32-.

Carbon Steel Corrosion in Simulated Anoxic Concrete Pore Water for Nuclear Waste Disposal Application

Carbon steel is a potential canister material for high-level waste disposal in underground geologic repository systems where reducing conditions dominate after the thermal period. For example, the super container concept in the proposed Boom Clay repository concept in Belgium comprises a carbon steel overpack containing the waste and an outer stainless steel liner forming the boundary of an annular region containing a cementitious material. The outer surface of the carbon steel overpack would be in contact with concrete pore water with pH near 12.5, expected to persist for thousands of years. This study investigated carbon steel passivity, corrosion rate, repassivation, hydrogen generation, and hydrogen induced cracking tendency in simulated concrete pore water under anoxic conditions, and the effects of possible anions in the pore water such as Cl−, , and S2−. Carbon steel maintained its passivity at 50°C and 80°C up to a test period of 6 months and passive corrosion rates were mostly in the range of 0.1...

Corrosion issues of carbon steel radioactive waste packages exposed to cementitious materials with respect to the Belgian supercontainer concept

ABSTRACTThe supercontainer is the reference concept for the post-conditioning of vitrified high-level radioactive waste and spent fuel in Belgium. It comprises a prefabricated concrete buffer that completely surrounds a carbon steel overpack. In this highly alkaline environment (pH ∼13.6) and under normal conditions (i.e. without the ingress of aggressive species), the carbon steel overpack will be protected by a passive oxide film, which is expected to result in very low uniform corrosion rates. Considering the long timescales applicable to waste disposal, determining accurate and reliable estimates of the uniform corrosion rate under anoxic conditions forms a very important part of the safety and feasibility strategy. On the other hand, the occurrence of localised forms of corrosion cannot be neglected and, therefore, it is also crucial to demonstrate that no other form of corrosion apart from uniform corrosion can take place. This paper gives an overview of the status of the research into the uniform corrosion, pitting corrosion and stress corrosion cracking behaviour of P355 QL2 grade carbon steel exposed to an artificial concrete pore water solution.This paper is part of a supplement on the 6th International Workshop on Long-Term Prediction of Corrosion Damage in Nuclear Waste Systems.

SCC susceptibility of carbon steel radioactive waste packages exposed to concrete porewater solutions under anoxic conditions

ABSTRACTThe supercontainer (SC) is the reference concept for the post-conditioning of vitrified high-level radioactive waste and spent fuel in Belgium. It comprises a prefabricated concrete buffer that completely surrounds a carbon steel overpack. Welding is being considered as a final closure technique of the carbon steel overpack in order to ensure its water tightness. Welding is known to induce residual stresses near the weld zone, which may lead to an increased susceptibility to stress corrosion cracking (SCC). In this study, slow strain rate tests were conducted to study the SCC behaviour of plain and welded P355 QL2 grade carbon steel exposed to an artificial concrete porewater solution. The tests were performed at 140°C, a constant strain rate of 5 × 10−7 s−1 and at open circuit potential under anoxic conditions. The effect of sulphide on the SCC behaviour was investigated up to levels of 15.6 mM S2− (500 mg L−1).This paper is part of a supplement on the 6th International Workshop on Long-Term Prediction of Corrosion Damage in Nuclear Waste Systems.

Summary of studies on the anaerobic corrosion of carbon steel in alkaline media in support of the Belgian supercontainer concept

ABSTRACTIn the Belgian supercontainer concept, a carbon steel overpack will surround high-level waste and spent fuel containers and be encased in a cementitious buffer material. A programme of research has been carried out to investigate and measure the rate of anaerobic corrosion of carbon steel in an artificial alkaline porewater that simulates the aqueous phase in the cementitious buffer material and also in solid cement matrices. This paper highlights the main features of the programme and presents some recent results from the analysis of several samples, including two 20-year-old steel in cement samples, which showed that under anoxic conditions there is no localised corrosion or incorporation of iron into the surrounding cement, even in the presence of 20 000 ppm chloride in the porewater.This paper is part of a supplement on the 6th International Workshop on Long-Term Prediction of Corrosion Damage in Nuclear Waste Systems.

Predictive modelling of the corrosion rate of carbon steel focusing on the effect of the precipitation of corrosion products

ABSTRACTIn the frame of conceptual design for the long-term management of high level radioactive waste defined by Andra (France), it is assumed that carbon steel (C-Steel) overpack would be mainly affected by general corrosion. As it is widely accepted that precipitation of a corrosion product layer (CPL) can limit the corrosion rate, the aim of this work is to simulate the growth of a thick FeCO3 CPL and its impact on the evolution of the corrosion rate. We use a 1D finite element method numerical model based on the resolution of Nernst–Planck equation in free potential conditions. Two numerical models of the evolution of the corrosion rate of low alloy steel in anoxic conditions are developed. These models take into account the growth of a porous CPL assuming either a constant and homogeneous porosity or a varying and heterogeneous porosity and results are compared to experimental data.This paper is part of a supplement on the 6th International Workshop on Long-Term Prediction of Corrosion Damage in Nuclear Waste Systems.

Re-evaluation of the required thickness of the carbon steel overpack for high-level radioactive waste disposal in Japan based on the latest scientific and engineering knowledge

ABSTRACTThe required thickness of the carbon steel overpack for high-level radioactive waste geological disposal in Japan was re-evaluated based on the latest scientific and engineering knowledge. The re-evaluation of the required thickness was carried out for the concept of vertical emplacement in a hard rock system (which requires the greatest overpack thickness). The re-evaluation was carried out to meet design requirements for corrosion resistance, structural integrity and the prevention/reduction of detrimental radiation effects on corrosion resistance. The re-evaluated minimum thicknesses to assure a lifetime greater than 1000 years were 122 mm for the flat end plate and 91 mm for the cylindrical body. Therefore, the applicability of the reference thickness (190 mm) for the overpack was reconfirmed.This paper is part of a supplement on the 6th International Workshop on Long-Term Prediction of Corrosion Damage in Nuclear Waste Systems.

Plutonium Migration in Compacted Bentonite with Iron Corrosion for 15 Years

ABSTRACTIn disposal of high-level radioactive waste, carbon steel overpack will be corroded after closure of the repository, creating a reducing, low-pH environment around the repository. A plutonium diffusion experiment was performed over 15 years with Kunigel V1, which is a typical Japanese bentonite that contains about 50% montmorillonite, in contact with an iron coupon. A tracer solution (10 µL) containing 1 kBq of 238Pu was applied at the interface between the iron coupon and compacted bentonite that was saturated with deionized water. After the diffusion period, the plutonium distribution in the bentonite specimen was measured with an alpha scintillation counter, and the iron and sodium distributions were obtained by inductively coupled plasma-mass spectrometry. Plutonium penetrated into the bentonite to a depth of 2 mm, and more than 90% of plutonium remained in corrosion product at the interface. The bentonite around the interface was dark green like green rust or magnetite according to visual observation. Iron was detected throughout the bentonite and there was a particularly high iron concentration from the interface to a depth of 2 mm, whereas the sodium concentration decreased slightly from the interface to 2 mm. We proposed that ferrous ions diffused into bentonite as the iron coupon corroded and precipitated such as magnetite. The magnetite precipitation would decrease the bentonite pH, resulting in the dissolution and migration of plutonium. Small cracks were also observed 2 to 6 mm from the interface and could have been formed by the migration of hydrogen generated by corrosion.