High-level Radioactive-waste Containers Research Articles

Geological analysis of the hydro-mechanical responses of Opalinus claystone to excavation-induced damage via engineering

In the realm of engineering geology, specifically concerning the containment of high-level radioactive waste, this study examines the excavation-induced damage zone (EDZ) in underground drifts and its evolution, which is critical for the effectiveness of underground repositories. The focus is on the self-sealing capability of the EDZ in Opalinus claystone (OPA) under mechanical compression and hydration induced by a bentonite core sealing component post-nuclear waste emplacement. This core applies a swelling pressure of 5 MPa to the drift walls during resaturation, affecting repository integrity. Using numerical modeling, this study pioneers the simulation of real-time self-sealing mechanisms in OPA, a crucial geological barrier in nuclear waste containment. The approach integrates a plastic damage model with an auxiliary deformation model and Biot’s model to accurately depict moisture-dependent OPA deformation. This modeling is essential for understanding claystone geomechanics in nuclear waste repositories. Detailed post-analysis of crack features and unsaturated fractured OPA permeability, following fracture energy regularization principles and the cubic law, provides insights into the geotechnical properties of the EDZ, enhancing the accuracy of geological forecasts in nuclear waste management. The model’s reliability is confirmed through various numerical simulations, including triaxial compression tests, observations of swelling deformations due to moisture variations, and comprehensive water permeability tests assessing the self-sealing efficiency of fractured OPA during water injection. In situ experiments at the Mont Terri Rock Laboratory support these findings, particularly regarding EDZ self-sealing under combined mechanical and hydration conditions. Comparative analysis of numerical results and experimental data validates the model’s ability to replicate fractured OPA self-sealing behavior, crucial for assessing nuclear waste repository safety and effectiveness. Effective self-sealing is underscored by reduced water permeability in the EDZ during resaturation, affirming the model’s relevance in evaluating nuclear waste containment strategies. Moreover, insights from this model extend to assessing CO2 migration, given claystones’ role as caprocks in CO2 geological sequestration schemes, thereby enhancing its importance in long-term environmental safety evaluations for both nuclear waste and CO2 storage scenarios.

Effect of temperature on mineral evolution and adsorption of Eu3+ by bentonite in hyperalkaline conditions

Abstract Compacted bentonite, used as an engineering barrier for permanent containment of high-level radioactive waste, is susceptible to mineral evolution resulting in compromise of the expected barrier performance due to alkaline–thermal chemical interaction in the near-field. To elucidate the mineral-evolution mechanisms within bentonite and the transformation of the nuclide adsorption properties during that period, experimental evolution of bentonite was conducted in a NaOH solution with a pH of 14 at temperatures ranging from 60 to 120°C. The results showed that temperature significantly affects the stability of minerals in bentonite under alkali conditions. The dissolution rate of fine-grained cristobalite in bentonite exceeds that of smectite, with the phase-transition products of smectite being temperature-dependent. As the temperature rises, smectite experiences a three-stage transformation: initially, at 60°C, the lattice structure thins due to the collapse of the octahedral sheets; at 80°C, the lattice disintegrates and reorganizes into a loose framework akin to albite; and by 100°C, it further reorganizes into a denser framework resembling analcime. The adsorption properties of bentonite exhibit a peak inflection point at 80°C, where the dissolution of the smectite lattice eliminates interlayer pores and exposes numerous polar or negatively charged sites which results in a decrease in specific surface area and an increase in cation exchange capacity and adsorption capacity of Eu3+. This research provides insights into the intricate evolution of bentonite minerals and the associated changes in radionuclide adsorption capacity, contributing to a better understanding of the stability of bentonite barriers and the effective long-term containment of nuclear waste.

Pilot plant tests to demonstrate the functionality of sealing elements made of salt cut bricks

Abstract. The long-term safe containment of high-level radioactive waste in a repository in rock salt is ensured if the geological barriers in conjunction with the geotechnical barriers are permanently impermeable to fluids. As such, an essential factor in underground disposal is to confirm that the interfaces between the biosphere and the lithosphere, i.e., shafts, boreholes, and galleries, created during the excavation of underground cavities can be sufficiently tightly sealed. An essential element of the sealing system required to this end is shaft closure. All shaft closure concepts developed thus far include sealing and supporting elements in repository shafts, but differ in the arrangement of these structural elements and the materials used. The materials currently proposed and planned for the construction of the sealing elements include: clay/bentonite, asphalt/bitumen, crushed salt, and salt/sorel cement. In addition to the materials mentioned above, a research project funded by the German Federal Ministry for Economic Affairs and Energy (Bundesministerium für Wirtschaft und Energie, BMWi) is investigating the possibility of integrating a layer of salt cut bricks several decameters to 100 m thick into the shaft closure system as a sealing element that provides the option of allowing the geological barrier to heal in the long term. Like the surrounding rock mass, the salt cut bricks are made of natural rock salt. According to this plan, the profile of the bricks is designed to minimize joint volume as far as possible by cutting them to match the geometry of the shaft. The joints between the salt cut bricks can either be filled with, for example, a supersaturated salt solution, Magnesium building materials, molten salt, crushed salt, etc., or directly brought into contact by wetting the surface of the salt cut bricks. Once the salt solution has hardened (cooling of the solution, evaporation of the mixing water), only the pore space in the crushed-salt joint sealant filled with salt solution or air, or the joint volume resulting from the mismatch between individual rock salt bricks are susceptible to a reaction. This means that the sealing element consisting of salt cut bricks develops early supporting pressure against the creeping rock salt of the rock mass compared to crushed salt, has a low initial porosity and already shows a strong sealing effect in the short term (regression of the loosened zone). One can also assume that cohesive bonding between the surrounding rock and the sealing element can already be achieved by introducing the joint filling or by wetting the contact surfaces of the salt cut bricks (no or reduced separation planes in the contact zone). Essential prerequisites for the investigation of the geomechanical-geohydraulic effectiveness of a sealing element made of salt cut bricks included the development and construction of a pilot plant to analyze the mechanical and hydraulic material properties of the bonding system comprising salt cut bricks and joint sealing (FKZ 02E11223, FKZ 02E11425), as well as preliminary investigations on the production of salt cut bricks and joint sealant (→ cutting technique/processing of salt cut bricks; maufacture/workability of jointing material) and on the spatial arrangement of the salt cut bricks (→ avoidance of continuous axial joints in the bonding system, Fig. 1). The presentation includes the results of the research work on the development, construction and commissioning of the pilot plant, as well as the first successful test results demonstrating the functionality of sealing elements made of salt cut bricks.

Long term corrosion estimation of carbon steel, titanium and its alloy in backfill material of compacted bentonite for nuclear waste repository

The container of high-level radioactive waste (HLRW) being in deep geological disposal, the backfill material is needed to serve as the second defense for HLRW and the highly compacted bentonite is generally selected. As the time goes, the underground water will infiltrate the backfill, causing the corrosion of materials for the building of containers in the formed electrolyte. Carbon steel, titanium and its alloy are the potential candidate materials for the fabrication of HLRW containers. The current investigation aims at assessing the safety of HLRW container in deep geological disposal for hundreds of thousands of years and facilitating the material selection for future container fabrication by estimating their corrosion behavior in compacted bentonite with a series of moisture content at different temperatures through electrochemical methods including open circuit potential (OCP), electrochemical impedance spectroscopy (EIS) and potentiodynamic polarization curve (PC) measurements. The corrosion rates were estimated for a carbon steel, a pure titanium and a titanium alloy in compacted Gaomiaozi Bentonite infiltrated with simulated underground water in Beishan area of China over an expected disposal period up to 106 years respectively, showing that titanium and its alloy are more reliable materials for building HLRW containers than carbon steel.

Experimental simulation of boundary condition effects on bentonite swelling in HLW repositories

On the basis of textural and hydro-mechanical characteristics, bentonite has been proven to be an effective buffer/backfill material for long-term containment of high-level radioactive waste (HLW) in deep geological repositories. Herein, the results of experiments performed to investigate the swelling equilibrium limit (SEL) of bentonite under various boundary conditions are presented. A special apparatus was employed to simulate various stress–strain boundary conditions, including constant volume (CV), constant vertical stress (CVS), and constant stiffness (CS). Bentonite samples were prepared with various initial dry densities ranging from 1.5 to 1.7 g/cm3 and vertically stressed to different levels. During wetting, they were subjected to different boundary conditions before the swelling strain or swelling pressure reached equilibrium. Test results indicate that stress–strain boundary conditions have significant effects on the measured swelling strain and swelling pressure of the tested bentonite. More specifically, the relationship between the sequence of swelling pressure under different boundary conditions is CV > CS > CVS, while the relationship between the sequence of swelling strain is CVS > CS > CV. In addition, the characteristics of SEL curves are governed by the initial dry density and vertical stress with the effect of dry density being more significant. Based on these results, several SEL curves were developed to index the effects of boundary conditions on swelling potential of bentonite. They can be used to evaluate the final stress and volume states of bentonite during fluid infiltration under the range of boundary conditions possible in HLW repositories.

Crevice corrosion of copper for radioactive waste packaging material in simulated groundwater

The investigation described here was conducted to clarify the corrosion behaviour of high level radioactive waste containers made of copper. The influences of oxygen, chloride ion and sulphate ion on copper crevice corrosion were studied in solutions simulating groundwater characteristic of northwest China. The results showed that oxygen, chloride ion and sulphate ion promote crevice corrosion. Chloride ion was found to play a significant role in the crevice corrosion mechanism in copper, but sulphate ion had no effect on the mechanism.

Tectonic Processes Modeling For High-Level Radioactive Waste Disposal

The possibility of using deep geological formations to dispose of high-level radioactive waste (HLW) is a subject raising heated debate among scientists. In Russia, the idea of constructing HLW repository in the Niznekansky granitoid massif (NKM) in Krasnoyarsk area is widely discussed. To solve this problem we are elaborating a technology associated with time – space stability prediction of the geological environment, which is subject to geodynamic processes evolutionary effects. It is based on the prediction of isolation properties stability in a structural tectonic block of the Earth’s crust for a given time. The danger is in the possibility that the selected structural block may be broken by new tectonic faults or movements on a passive fault may be activated and thus underground water may penetrate to HLW containers.

Corrosion of High Purity Copper in Solutions Containing NaCl, Na2SO4 and NaHCO3 at Different Temperatures

Oxygen free copper is one of the candidate materials for the design of high level radioactive waste containers, due to its excellent resistance to generalized and localized corrosion in aqueous electrolytes; particularly in reducing environments as the ones expected in the repository site.The aim of this work is to contribute to determine the material durability from the corrosion point of view. To this purpose the anodic behavior of copper was studied in solutions with different electrolytes of interest for being present in groundwaters that might be in contact with the containers. The research program began with the study of corrosion susceptibility in chloride solutions, one of the main aggressive species to corrosion, and continued with the study of the behavior of copper in solutions with sulfate and bicarbonate, which are also present in the groundwaters that might be in contact with the repository. The corrosion potential was monitored and anodic polarization curves were performed varying the temperature and the concentration of ions in deaerated solutions of chloride, sulfate and bicarbonate. The study was complemented with microscopic observations of the corrosive attack and compositional determinations using Energy Dispersive Microanalysis (EDS).Preliminary results showed the aggressiveness of chloride ions: the corrosion potential decreased with the ion concentration and with temperature. The current density increased both with temperature and with chloride concentration. A breakdown potential was observed under certain conditions. Corrosion products showed a composition of 30 wt% Cl and 70 wt% Cu. In sulfate containing solutions, the corrosion potential decreased with temperature but increased slightly with the ion concentration, whereas the current density increased with temperature and it was not significantly affected by the concentration. A breakdown potential was identified, which decreased with temperature and increased with sulfate concentration. In bicarbonate containing solutions, the corrosion potential decreased clearly with temperature and slightly with concentration. Different characteristics were observed in the curves: current densities were lower than in sulfate containing solutions and an anodic peak appeared under certain conditions.

First corrosion stages in Tournemire claystone/steel interaction: In situ experiment and modelling approach

Steels are possible materials for high-level radioactive waste containers used in long-term geological disposal in claystone environments. The French Institute for Radioprotection and Nuclear Safety (IRSN) is conducting an experimental programme on claystone/steel interactions in underground natural conditions at its Tournemire Underground Research Laboratory (URL, Aveyron, France). In this programme, several experimental conditions have been tested: i) the location in the geological formation, ii) the nature of the steel, iii) the claystone texture, and iv) the water content. This programme has investigated, after six years of experimentation, the effect of these parameters on the degree of steel corrosion, the spatial transport of iron and the claystone/steel interaction processes. Whatever the context results indicate corrosion rates of about 3μm/year for carbon steels and about 0.1μm/year for stainless steels. A detailed petrological study of the claystone/steel interface reveals a clear influence of free water content and claystone texture on the iron transport processes. All the data obtained from this study demonstrate that, during the initial oxic stage of steel corrosion, oxygen entrapped in the claystone porosity reacts with iron to form goethite/lepidocrocite minerals. Even after six years of reaction, there is no observed evidence of the anoxic corrosion of steel, irrespective of the context. The precipitation of iron hydroxide within the claystone causes intense calcite dissolution and a slight dissolution of mixed-layered illite–smectite. The minimum amount of oxygen required to form goethite in our experiments is estimated by thermodynamic simulations. This amount of oxygen is nearly the same as the amount of iron corroded. Moreover, additional simulations show that siderite forms when goethite becomes unstable under anoxic condition, and that the Tournemire mineral assemblage buffers the pore-water pH at a value close to neutrality, which prevents magnetite formation. Conclusions underline the important role played by oxygen entrapped in the near-field formation on the first iron corrosion stage. The quantification of the amount of oxygen present in the near field is a critical parameter that has to be taken into account in the nuclear waste storage safety assessment.

Mineralogical reactions in the Tournemire argillite after in-situ interaction with steels

Steels are possible materials for high level radioactive waste containers used in long term geological disposal in argillaceous environments. Under chemical conditions of a deep repository, the release of iron from these canisters could modify the properties of clay minerals. Thereby, many batch experiments have been realized on argillite under disposal conditions. The French Institute of Radioprotection and Nuclear Safety conducts an experimental program on steel/argillite interactions in underground natural conditions in its Tournemire experimental site (Aveyron, France). In order to determine these interaction processes, the present study focuses both on petrography and mineralogy characterization of modifications in the argillite in contact with steel samples. The accurate study of the argillite and steels after 6 years of contact indicates no significant changes for the argillite in contact with the stainless steels. For the argillite in contact with the carbon steels both numerous steel corrosion evidences and clear textural, petrographic, mineralogical and chemical modifications of the initial nature of the argillite are observed. The Fe°/argillite contacts indicate the development of a Fe-rich front within the argillite resulting from the iron diffusion from the corroded steel and which is connected with mineralogical changes of the initial argillite. This Fe-enrichment is accompanied by crystallization of significant amounts of goethite/lepidocrocite (FeOOH) and some traces of magnetite near the steel surface. The formation of Fe-rich phases within the argillite is associated with dissolution–crystallization processes. Thus, after bulk sample XRD and SEM data, these processes consist in dissolution of calcite, phyllosilicate (I/S) and pyrite and crystallization of some traces of gypsum and melanterite.

PREDICTION OF THE TECTONIC PROCESSES AT A SELECTED AREAS FOR DISPOSE RADIOACTIVE WASTE

The possibility of using deep geological formations to dispose of high-level radioactive waste (HLW) is a subject raising heated debate among scientists. In Russia, the idea of constructing HLW repository in the Niznekansky granitoid massif (NKM) in Krasnoyarsk area is widely discussed. To solve this problem we are elaborating a technology associated with time - space stability prediction of the geological environment, which is subject to geodynamic processes evolutionary effects. It is based on the prediction of isolation properties stability in a structural tectonic block of the Earth’s crust for a given time. The danger is in the possibility that the selected structural block may be broken by new tectonic faults or movements on a passive fault may be activated and thus underground water may penetrate to HLW containers.

Degradation of phosphatic waste forms incorporating long-lived radioactive isotopes

AbstractThis paper provides a brief perspective on synthetic, phosphate-based waste forms for high level radioactive waste (HLW). Evidence in support of their long-term stability is then discussed by reference to the degradation of natural monazites with emphasis on the fate of released uranium, thorium and the rare earths (REE). It is apparent that the REE can be mobilized and fractionated at temperatures anticipated in a HLW repository (∼200°C). This provides an indication of the likely fate of the trivalent actinides (Am(III), Cm(III)) if incorporated in similar matrices. Thorium, though released on alteration of monazite, tends to re-concentrate locally in secondary, microcrystalline phases. In relative terms, U is readily removed from monazites. Although it can be re-concentrated in alteration products, the potential exists for substantial loss of U to groundwater. The findings of this research have important implications for the performance of radioactive waste disposal systems where there is a clear need for improved chemical data to describe the precipitation-dissolution of phosphate phases. It is concluded that monazite-like ceramics designed for the containment of HLW will retain tetravalent actinides but may release uranium in response to natural degradative processes.

環境有害物質処理プロセスと装置材料 高レベル放射性廃棄物の地層処分容器材料の腐食寿命評価

環境有害物質処理プロセスと装置材料 高レベル放射性廃棄物の地層処分容器材料の腐食寿命評価

Performance assessment of multiple engineered barrier systems

The lifetime of the engineered barrier system used for containment of high-level radioactive waste will significantly impact the total performance of a geological repository facility. Currently two types of designs are under consideration for an engineered barrier system, single engineered barrier system and multiple engineered barrier system. Multiple engineered barrier system consists of several metal barriers and the waste form (cladding). Some recent work showed that a significant improvement of performance can be achieved by utilizing multiple engineered barrier systems. Considering sequential failures for each barrier, we model the reliability of the multiple engineered barrier system. Weibull and exponential lifetime distributions are used through out the analysis. Furthermore, the number of failed engineered barrier systems in a repository at a given time is modeled using a Poisson approximation.

A Phenomenological Approach to Simulating the Evolution of Radioactive-Waste Container Damage Due to Pitting Corrosion

AbstractThe damage to high-level radioactive-waste containers by pitting corrosion is an important design and performance assessment consideration. It is desirable to calculate the evolution of the pit depth distribution, not just the time required for initial penetration of the containers, so that the area available for release of radionuclides through the container can be estimated. A phenomenological approach for computing the time evolution of these distributions is presented which combines elements of the deterministic and stochastic aspects of pit growth. The consistency of this approach with the mechanisms believed to control the evolution of the pit depth distribution is discussed. Qualitative comparisons of preliminary model predictions with a variety of experimental data from the literature are shown to be generally favorable. The sensitivity of the simulated distributions to changes in the input parameters is discussed. Finally, the results of the current model are compared to those of existing approaches based on extremevalue statistics, particularly regarding the extrapolation of laboratory data to large exposed surface areas.

Reliability modeling of engineered barrier systems for nuclear waste: A conditional approach

The lifetime of the Engineered Barrier System (EBS) used for containment of high-level radioactive waste will play an important role in controlling the cumulative release of radionuclides from a geological repository. Essentially, an EBS has two major components: a metal barrier and the waste form (cladding). Since any suitable site is above the static water table, the decay heat generated from this waste is expected to keep the surface temperature of a canister above the boiling point of water for a T random time period. Assuming the Weibull lifetime distribution on each component, we use the conditional approach to derive the lifetime distribution of the EBS. In our model, the oxidization process of the metal barrier (i.e. the lifetime of the metal barrier) starts after this time T and the lifetime of the cladding starts after the lifetime of the metal barrier

Kinetic model calculations of colloid growth in NaCl

The formation of radiation damage in NaCl has generated interest because of the relevance of this damage to the disposal of radioactive waste in rock salt formations. In order to estimate the long-term behaviour of rock salt under irradiation, an accurate theory describing the major processes of radiation damage in NaCl is required. In 1977 Jain and Lidiard presented a rather simple model which, until now, has served as a means for long-term predictions of near field radiation damage processes in rock salt repositories. Although the model of Jain and Lidiard describes many experimental results on colloid growth in NaCl rather well, it neglects the colloid nucleation stage and the effect of impurities on colloid growth. The model presented in this paper is an extended version of the Jain-Lidiard model; its extensions comprise the effect of impurities and the colloid nucleation stage on the formation of radiation damage. The extended model has been tested against various experimental data obtained from the literature and accounts for several well known aspects of radiation damage in alkali halides which were not covered by the original Jain-Lidiard model. Using this extended model, we have calculated the amounts of NaCl that can be converted into metallic Na and molecular Cl2 for various Dutch concepts for reprocessed high-level radioactive waste (HLW) repositories. The theory predicts that the concentration of these defect aggregates in rock salt, even very close to the HLW containers, will be limited to a few mol%.

Stochastic Modeling of the Influence of Environment on Pitting Corrosion Damage of Radioactive-Waste Containers

AbstractA physically-based, phenomenological stochastic model for pit initiation and growth is presented as a potential tool for predicting the degradation of high-level radioactive-waste containers by aqueous pitting corrosion. Included in the model are simple phenomenological equations describing the dependence of the controlling stochastic parameters on the applied (or corrosion) potential, chloride ion concentration, and absolute temperature. Results from this model are presented that demonstrate its ability to simulate several important environmental effects on pitting.

Modeling pitting corrosion damage of high-level radioactive-waste containers using a stochastic approach

Stochastic models of pitting corrosion have been explored and found to be potentially useful in predicting damage of high-level radioactive-waste containers. These computer models include simple phenomenological relationships describing the environmental dependence of the stochastic parameters, and can simulate pit initiation and growth under various environments, including those that change during exposure. Equally important, the models provide quantitative information that would be useful to performance assessment codes for the entire waste repository system. Valuable insights into the mechanisms of pit initiation and growth also have resulted from this research. Finally, comparisons between the stochastic and mechanistic approaches to modeling pitting, and directions for future research are discussed.

Corrosion of Stainless Steel for HLW Containers Under Gamma Irradiation

ABSTRACTThe corrosion behavior of type 304 stainless steel was studied under gamma irradiation as part of the evaluation for the long-term durability of high-level radioactive waste (HLW) disposal containers. Gamma rays, generated from fission products in high-level radioactive waste, are considered to change the environment around the canisters and overpacks. The redox potentials for NaCl solutions and corrosion potentials of stainless steel were measured to consider the effects of gamma irradiation, by using an electrochemical method. The pitting potentials of stainless steel for NaCl solutions were also measured to examine the pitting corrosion under gamma irradiation. As a result of this experiment, it is concluded that the oxidizing properties as a result of the formation of H2O2 and H2 produced by gamma irradiation depended on the concentration of Cl−, and that the strength of oxidizing properties of 1M (mol·dm−3) NaCl solution was particularly high. and the pitting corrosion was found for 1M NaCI solution under gamma irradiation at the dose rate of 2.6×102 C/kg·h (1.0×106 R/h) at 60°C, by using an electrochemical method.