Disposal Of High-level Nuclear Waste Research Articles

Machine learning-based models for predicting gas breakthrough pressure of porous media with low/ultra-low permeability.

Gas breakthrough pressure is a significant parameter for the gas exploration and safety evaluation of engineering barrier systems in the carbon dioxide storage, remediation of contaminated sites, and deep geological repository for disposal of high-level nuclear waste, etc. Test for determining gas breakthrough pressure is very difficult and time-consuming, due to the low/ultra-low conductivity of the specimen. It is also difficult to get a comprehensive and high-precision model based on limited results obtained through individual experiments, as the measurements of gas breakthrough pressure were influenced by many factors. In this study, a collected database was built that covered a lot of former test data, and then, two models were developed by the random forest (RF) algorithm and multiexpression programming (MEP) method. The MEP model constructed with explicit expressions for the gas breakthrough pressure overcame the drawbacks of common "black box" models. Meanwhile, five significant indicators were selected from ten common features using the permutation importance algorithm. The RF model was interpreted by the Shapley value and the PDP/ICE plots, while the MEP model was analyzed through the proposed explicit expression, showing strong consistence with that in former studies. Finally, robustness analysis was conducted, and stability of the proposed two models was verified.

Controlled disorder for the Yb2Ti2-xO7-2x (x = 0 to 1) series and corresponding radiation tolerance

The use of ceramics as radionuclide containment matrices for the safe long-term storage and disposal of high-level nuclear waste is studied with emphasis on chemistry flexibility, structural stability, and radiation tolerance. Here we investigate the flexibility of a particular series of compounds Yb2Ti2-xO7-2x (x = 0, 0.2, 0.52, 0.8, and 1) to form a solid solution between the end members Yb2Ti2O7 of pyrochlore structure through to Yb2TiO5 of fluorite structure. Whilst it is shown that cation disorder and corresponding transition from the ordered pyrochlore to disordered fluorite is facilitated via increasing the ytterbium to titanium ratio, it is also shown that sintering conditions may be controlled in such a way as to encourage the growth of the pyrochlore structure even in Yb2TiO5. The radiation response of these materials was tested in situ via 1 MeV krypton irradiation coupled with transmission electron microscopy characterisation. The critical fluence of ions required to transition the crystalline phase to amorphous was found to increase with increasing disorder from Yb2Ti2O7 to Yb2TiO5. However, of the two forms (pyrochlore and fluorite) of Yb2TiO5 studied via ion irradiation that with the greater pyrochlore structure and so more ordered also had the lower critical temperature, 434 K, for maintaining crystallinity during irradiation.

Enhanced Crevice Corrosion of Stainless Steel 316 By Degradation of Cr-Containing Hollandite Crevice Former

The disposal of high-level nuclear waste (HLW) is an important societal problem. It is also an extremely challenging corrosion research topic due to the highly complicated waste chemistry, long lasting radiation effect, and inevitable exposure to aqueous environments for hundreds of thousands of years. To safely accommodate some critical radionuclides such as volatile 129I and strongly heat generating 137Cs, numerous crystalline ceramic waste forms have been created and studied. Among the promising nuclear waste forms, hollandite was developed to isolate Cs waste and the corresponding decay product Ba due to the open tunnel structure capable of accommodating mono- or divalent cations. In this study, the synergistic corrosion interactions between a Cr-containing hollandite (Ba1.15Cr2.3Ti5.7O16) and stainless steel (SS) 316 is explored. This is relevant to the permanent disposal of HLW involving the encasement of glass or crystalline ceramic waste forms containing immobilized radionuclides in a metallic canister made from corrosion resistant alloys such as SS. The experiments performed in this study simulate the potential corrosion interactions occurring at the interface of the SS and ceramic. After corroding the SS316 and Cr-containing hollandite in proximity in 0.6 M NaCl at 90 oC for 28 days, severe crevice corrosion was identified on the surface of the SS, as evidenced by the presence of large pits and crevice damage with diameters of hundreds of microns. Similarly, localized damage was also found at matching sites on the Cr-hollandite surface, indicating interactions between the two materials. The synergistic corrosion interaction was likely due to the continuous release of Cr3+ cations from both materials, including the passive dissolution of the SS and the heterogenous degradation of the Cr-hollandite. The extra Cr3+ cations originated from the hollandite reduces the incubation time required to reach the critical crevice solution condition, thereby accelerating the breakdown of the passive film of SS and the subsequent onset of active dissolution. An adapted crevice corrosion model is applied to quantitively explain the accelerated corrosion of SS observed in this study.

Effect of thermal ageing on physical properties of MX80 bentonite under high-temperature conditions

In a deep geological repository (DGR) system, the buffer layer is indispensable to ensuring the safe disposal of high-level radioactive nuclear waste (HLW). Because the heat generated by the decay of the spent nuclear fuel in a canister is released to the surrounding buffer layers, the bentonite buffer material experiences long-term high-temperature conditions. Therefore, the variations in physical properties of bentonite buffer material under high-temperature conditions are one of the important parameters in the DGRs for HLW. A series of tests on the specific gravity, specific surface area (SSA), Atterberg's limits, and free swelling ratio of MX80 bentonite after heating for different times at a high temperature of 200 °C were conducted, to investigate the influence of thermal ageing time on its physical properties. Then, the microscopic investigations, including X-ray diffraction (XRD), thermogravimetric analysis (TGA), and scanning electron microscopy (SEM), were conducted to explain the aforementioned variations from the microscopic point of view. Based on the crystal layer structure characteristics of montmorillonite, the quantitative relationships between the physical properties of bentonite and its mineral composition or bound water content were established to further explain the mechanism by which thermal ageing affects the physical properties of MX80 bentonite. The results indicate that with increasing heating time, the specific gravity, SSA, liquid limit, plastic limit, plastic index and free swelling ratio decrease sharply by 2.5%, 4.5%, 3.1%, 5.5%, 2.8% and 30.2%, respectively, within 15–30 days of heating. After 30 days of heating, their variations are negligible. Under high-temperature conditions, the transformation of mineral composition, desorption of bound water, and changes in the micro-morphology are the fundamental reasons for the variations in the physical properties of bentonite, and they influence and interact with each other.

Evaluation of hydro-mechano-chemical behaviour of bentonite-sand mixtures

Bentonite-sand mixtures are widely used in engineering barrier of deep geological disposal of high-level radioactive nuclear waste and anti-seepage barrier of civil geotechnical engineering. Under the action of groundwater solution infiltration and external stress, the hydro-mechanical (HM) behaviour of bentonite-sand mixtures, i.e. the swelling characteristics and permeability, will change. Once the anti-seepage and filtration effect is weakened or lost, the pollutants will spread to the biosphere. Therefore, it is necessary to study the swelling characteristics and permeability of bentonite-sand mixtures under coupled mechano-chemical (MC) effect and to establish corresponding prediction model. For this reason, swelling tests under salt solution with different concentrations are conducted on pure bentonite and its mixtures with 30%, 70% and 90% sand contents, the compression tests are carried out on saturated samples, and the saturated permeability coefficient k of the sample under each load is calculated by Terzaghi's one-dimensional consolidation theory. The concepts of true effective stress p e , montmorillonite void ratio e m and critical sand content α s are introduced to determine the e m - p e relationship and finally the k - e m relationship of bentonite-sand mixtures. It is found that when the sand content α ≤ α s , the e m - p e relationship of the mixture is linear and independent of the salt solution concentration, and when α > α s , the e m - p e relationship of bentonite-sand mixture is bi-linear with the true effective deviatoric stress p esα as the intersection. In addition, the e m - k relationship also shows the linear trend when α ≤ α s , and the slope of the line increases with the increase of the salt solution concentration. When α > α s , the k - e m relationship will deviate from the linear relationship. Moreover, the larger the sand content is, the farther the deviation is. On the basis of summing the regularity, a model for predicting the HM behaviour of bentonite-sand mixture under the coupled MC effect is proposed. By comparing the swelling and permeability test results with model prediction results of different types of bentonite and its sand mixtures, the predictive model is verified. The study on the HM behaviour of bentonite-sand mixtures under salt solution infiltration and the model establishment can provide experimental and theoretical basis for the design and construction of anti-seepage engineering by bentonite-sand mixtures.

Hydrogen Permeation Behavior of Carbon Steel During Corrosion in Highly Pressed Saturated Bentonite

Deep geological disposal is the most reliable method for high-level nuclear waste, of which metal container as the first barrier for deep geological disposal of high-level nuclear waste is particularly important. Carbon steel is used as a container material because of the low possibility of local corrosion in bentonite. However, after the storage is closed, the decrease of oxygen content will create a near-field environment where the hydrogen embrittlement (HE) in the corrosion process of the container could happen. To evaluate the safety of containers in deep geological disposal of Beshai, the preselected area in China, hydrogen permeation efficiency and HE were estimated in highly pressed saturated bentonite by electrochemical and extrapolation analyses. It is concluded that hydrogen permeation efficiency increases with the disposal year, which proves that the hydrogen evolution reaction dominates the cathode process in the corrosion during long-term disposal. However, slow strain rate tensile shows that Q235 steel has a low HE sensitivity.

Discrete fracture network (DFN) modelling of a high-level radioactive waste repository host rock and the effects on its hydrogeological behaviour

Fracture network modelling and a hydrological evaluation were performed in a well more than 900 m deep that penetrated the Boda Claystone Formation, a potential host rock for high-level nuclear waste disposal facilities in Hungary. The fracture network geometry was generated with a discrete fracture network algorithm, in which the permeability and porosity of the system can be calculated if the aperture of the fractures is known. The hydrological aperture of the fractures was estimated via an aperture calibration based on a comparison of the measured and modelled permeability values. Flow zone indices were calculated for numerous sections along the well, designating hydraulic units, in which the fluid flow-controlling properties are internally even.Based on the fracture network geometry and the hydraulic flow units, most parts of the well behave uniformly, while three narrow zones differ significantly. The first zone is located in the upper 100 m of the well probably formed due to weathering. The second zone is located at approximately 400 m, where a large-scale structural boundary is presumed. In the third zone at 700 m, a lithological change greatly affects the hydrological properties, but the influence of tectonic processes cannot be ruled out.

The impact of 40 years of radiation on the integrity of copper

Cu piping previously used to carry process air was harvested from the now shutdown National Research Universal reactor at the Canadian Nuclear Laboratories site in Chalk River, Ontario. Metallurgical and microscopy examinations found no quantifiable differences between irradiated and non-irradiated sections of the pipe. This provides evidence for the safety case for the use of the Cu-coated used fuel container proposed by the Nuclear Waste Management Organization for the permanent disposal of high-level nuclear waste.

Development of triaxial compressive apparatus for neutron experiments with rocks.

Underground engineering for processes such as geological disposal of high-level nuclear waste, CO2 capture and storage, and mining and drilling for resources requires an understanding of the mechanical behavior of rocks at subsurface stress states, i.e., triaxial compressive stress. Strain measurement using neutron diffraction can be applied to rocks to analyze strain accumulation mechanisms at the microscopic scale. This study reports the development of triaxial compressive apparatus for strain measurement using neutron diffraction. The apparatus can analyze rock specimens (diameter, 25mm; length, 50mm) and apply a maximum confining pressure of 50 MPa. Materials for the components of the apparatus were investigated theoretically based on neutron beam transmission and experimentally using neutron diffraction experiments. The feasibility of the apparatus was verified by measuring strain at hydrostatic pressure under the application of confining pressure and triaxial compression. The theoretical and experimental results show that the apparatus could obtain sufficient neutron statistics from a rock specimen. It was confirmed experimentally that the measured strain values are correlated with the applied confining pressure and stress. The lattice strains of quartz minerals measured by neutron diffraction showed linear deformation behavior, indicating that elastic strain accumulated in the minerals. This apparatus will enable the finding of new insights into the deformation mechanisms of rocks.

The role of trust and risk perception in current German nuclear waste management.

One of the lessons learned in various countries that have to deal with spent nuclear fuel is that finding a proper place and siting a repository for high-level nuclear waste (HLW) cannot be achieved without public consent. After decades of obstruction, Germany recently launched a new, participatory, site-selection process for the disposal of HLW in deep geological formations. Nonetheless, significant opposition is assumed. Therefore, citizens' trust in the procedure and the agents involved may be paramount. We conducted an online survey (N ≈ 5000) in March/April 2020 to test a theoretical model on trust, perceived risks and benefits, and acceptance. We differentiated acceptance as a dependent variable according to distinct phases: the procedure, a possible decision on a disposal location, and the repository facility itself. The results show that trust is mainly important for explaining acceptance of the ongoing procedure and less so for the acceptance of the decision or the repository facility itself. Moreover, our investigation of the sample using a cluster analysis reveals characteristic patterns of trust, risk perception, and acceptance by three clusters: a cluster focusing on risk perception, an ambivalent cluster, and an indifferent cluster. Trust is lowest in the risk-focused cluster and highest in the ambivalent cluster.

Thermoplastic cohesive fracturing model of thermally-treated granite

Thermoplastic fracturing is a prominent characteristic of thermally-treated granite for temperature-dependent underground geotechnical engineerings , such as geothermal energy exploration and high-level nuclear waste disposal. More specifically, the tension-open of the cohesive crack in the fracture process zone (FPZ) presents a plastic softening feature dependent on temperature. However, understanding granite thermoplastic fracturing remains challenging. This work proposes a thermoplastic cohesive fracturing model of thermally-treated granite, incorporating temperature-dependent softening rule and completely coupled stress-strain-temperature constitutive relation. Several new thermoplastic fracturing properties in the proposed model are determined by further analyses of published fracturing test data (Miao et al., 2020) done on Beishan granite suffering high-temperature treatments (200–500 °C). The initial yield parameter (cohesive tensile strength) decreases linearly with the rising high-temperature treatments. The critical yield parameter (the critical crack opening displacement (COD)) follows a two-stage linear increasing rule with 300 °C as a boundary, similar to the temperature-dependent FPZ length. The plastic modulus, delineating the cohesive crack's softening rule, increases quadratically with the enhanced high-temperature treatment. The temperature sensitivity modulus ( T ), characterizing the contraction/expansion of yield surface with increasing temperatures, decreases from positive to negative with the increasing COD. Both the fracture energy and the accumulated dissipated energy present remarkable nonlinearities with several inflection points , implying the complex temperature-dependent fracture resistance . This proposed model was validated with three-point-bending fracturing tests of thermally-treated Beishan granite, fitting well (coefficient: 88.8%–99.7%) with the experimental cohesive tensile strength , critical COD, and the accumulated dissipated energy. Besides, by analyzing the softening curves of cohesive cracks in thermally-treated granite, the positive-negative change of T was consistent with the contraction-expansion transition of the yield surface. This work provides a modeling approach for temperature-dependent fracturing of rock-like materials. • The thermoplastic softening of FPZ is characterized via a thermoplastic framework. • This work proposes a thermoplastic cohesive fracturing model for rocks. • New model parameters were proposed to characterize temperature-dependent fracturing. • The correlation coefficient between the model and test data reaches 88.8%–99.7%. • This work provides a modeling approach for temperature-dependent fracturing of rock.

Michigan International Copper Analogue (MICA) project – current status

Abstract. One of the key requirements for the deep geological disposal of high-level nuclear waste is the assessment of its long-term performance and safety (up to 1 Ma). Regarding engineered barrier system materials, such as copper, much of the data available comes from short-term investigations, such as laboratory experiments at different scales. Copper is an important part of many waste packaging and disposal concepts, e.g. KBS-3 developed in Sweden and Finland and Mark II developed in Canada. Natural analogues provide another important way of obtaining understanding on potential repository system behavior. Observations made from the geological systems can be utilized in the safety case, providing information on the assessment time scale. Copper analogue studies (both natural analogues and archaeological analogues) have been reported in the literature and they have been extensively reviewed by various authors (e.g. Miller et al., 2000) and by safety case projects (e.g. Reijonen et al., 2015) within waste management organizations. So far, only a few studies have focussed on the general stability of native copper within its natural media (e.g. Milodowski et al., 2000; Marcos, 2002). Keweenaw native copper occurrences (Lake Superior, USA) have been mentioned as a qualitative source of information (e.g. in Miller et al., 2000); however, data to be used in process-based safety assessments for geological disposal are lacking. These deposits have been mined for a long time and there is a great deal of knowledge related to them as well as samples collected, but no formal review has been made from the geological disposal point of view. The native copper at the Keweenaw area reflects various geological environments from bedrock to sediment and even anthropogenic mine site remnants and geochemical environments (e.g., anoxic vs. oxic, sulphur-free vs. sulphur-bearing). It thus provides a unique complementary data source that will be useful for estimating processes governing behavior of metallic copper. The MICA project phase I systematically collects and reviews the existing literature and data on the Michigan copper analogue sites and available sampling potential. Here, we present the current status of the project.

A citizens workgroup helps researchers reflect on their work

Abstract. In our transdisciplinary project (http://www.transens.de) several academic disciplines work on questions and solutions for high-level nuclear waste disposal in Germany. Adding to this interdisciplinary setting, in our sub-project on trust, we have enlisted a group of 16 citizens (citizens workgroup, CWG), reflecting with us on our research with regard to research questions and approaches. In this talk, we present results from a joint workshop of researchers and the CWG on the role of trust in scientists. We want to understand how participants perceive and relate to science and how this may affect trust in scientists and experts. During a first workshop (conducted online in March 2021), the CWG members were allocated to three breakout groups (1)–(3) to discuss three guiding questions: a. What expectations do you have of a scientist from Social Sciences and Humanities (SSH)/Natural Sciences and Engineering (NSE) in the TRANSENS project? b. What characterizes a scientist from SSH/NSE that you trust? c. What would have to happen for you to no longer trust him/her? The group in breakout room (1) was asked to answer questions (a)–(c) for a fictitious SSH researcher. The group in breakout room (2) was asked to do the same for a fictitious NSE researcher. The group in breakout room (3) was asked whether they perceived differences in the trust of SSH and NSE researchers. For SSH we found that a scientist should be sympathetic and non-condescending, represent a neutral point of view, and consider all opinions. Remarkably, discussants in this group struggled to define a role for SSH scientists in high-level waste disposal research. Some participants ascribed SSH a moderating or mediating role. If necessary (e.g. communication of results), mediation between the NSE and the public can be added. SSH scientists may train other scientists with regard to their performance. Participants stated that scientists from NSE should present information in full and clarify the current state of research (provisional nature of knowledge acknowledged). A scientist should not conceal any information, uncertainties, or risks, be neutral and objective, have experience in the field (professionalism, experience, reputation), and not exclusively reproduce one's own opinion or spread untruths. In all groups, participants judged the NSE as “harder”, more serious and more tangible. It was also discussed that the scientist's appearance is of great importance and that a scientist can “pick up” the audience in an exciting way when communicating. Personal experience was mentioned several times in all groups as a basis for trust. These results match findings on the effect of value similarity (Siegrist et al., 2000, and own survey in 2020 in Germany): If a person is perceived as advocating similar values, it is more likely that I trust that person. Personal experience of – among other issues – similar values may increase trust. Moreover, and importantly, trust also emerges when participants know exactly how issues will be considered, e.g. if input from the CWG is considered or not – and if not: why? In general, participants want to be taken seriously. Thus, transparent and binding rules for all participants may be key for a relationship of trust.

Corrosion-Enhancing and Corrosion-Reducing Accessories in Bentonite Surrounding Copper-Shielded Containers for Nuclear Waste

Abstract The standard design for the underground disposal of high-level nuclear waste consists of a carbon steel container that hosts the waste shielded by a 5-cm-thick layer of corrosion-resistant...

Experimental determination of solubility constants of saponite at elevated temperatures in high ionic strength solutions

Abstract Saponite occurs in a wide range of environments from hydrothermal systems on the Earth to surface deposits on Mars. Of practical importance is that Mg-saponite forms when glasses for nuclear waste are altered in Mg-bearing aqueous solutions. In addition, saponite is favorably considered as candidate buffer material for the disposal of high-level nuclear waste and spent nuclear fuel in harsh environments. However, the thermodynamic properties, especially for Mg-saponites, are not well known. Here the author synthesized Mg-saponite (with nitrate cancrinite) following a previously reported procedure and performed solubility experiments at 80 °C to quantify the thermodynamic stability of this tri-octahedral smectite in the presence of nitrate cancrinite. Then, in combination with the equilibrium constant at 80 °C for the dissolution reaction of nitrate cancrinite from the literature, the author determined the solubility constant of saponite at 80 °C based on the solution chemistry for the equilibrium between saponite and nitrate cancrinite, approaching equilibrium from the direction of supersaturation, with an equilibrium constant of –69.24 ± 2.08 (2σ) for dissolution of saponite at 80 °C. Furthermore, the author extrapolated the equilibrium constant at 80 °C to other temperatures (i.e., 50, 60, 70, 90, and 100 °C) using the one-term isocoulombic method. These equilibrium constants are expected to have applications in numerous fields. For instance, according to the extrapolated solubility constant of saponite at 50 and 90 °C, the author calculated the saturation indexes with regard to saponite for the solution chemistry from glass corrosion experiments at 50 and 90 °C from the literature. The results are in close agreement with the experimental data. This example demonstrates that the equilibrium constants determined in this study can be used for reliable modeling of the solution chemistry of glass corrosion experiments.

Recent Advances in Corrosion Science Applicable To Disposal of High-Level Nuclear Waste.

High-level radioactive waste is accumulating at temporary storage locations around the world and will eventually be placed in deep geological repositories. The waste forms and containers will be constructed from glass, crystalline ceramic, and metallic materials, which will eventually come into contact with water, considering that the period of performance required to allow sufficient decay of dangerous radionuclides is on the order of 105-106 years. Corrosion of the containers and waste forms in the aqueous repository environment is therefore a concern. This Review describes the recent advances of the field of materials corrosion that are relevant to fundamental materials science issues associated with the long-term performance assessment and the design of materials with improved performance, where performance is defined as resistance to aqueous corrosion. Glass, crystalline ceramics, and metals are discussed separately, and the near-field interactions of these different material classes are also briefly addressed. Finally, recommendations for future directions of study are provided.

Swelling pressure of compacted MX80 bentonite/sand mixture prepared by different methods

Compacted bentonite/sand mixture have been considered as possible sealing/backfilling material in the deep geological disposal of high-level radioactive nuclear waste. The swelling pressure of the compacted bentonite/sand mixture is of significance in the design of the geological repository, which requires good consistency between data obtained by laboratory and field measurement. In this work, a series of swelling pressure tests were performed on compacted MX80 bentonite/sand specimens prepared by methods commonly adopted in laboratory (As-compacted, Transferred) and those mimicking the real block manufacturing process (Trimmed, Inserted). Results shown that with identical dry density (especially when the dry density was larger than 1.70 Mg/m3), largest swelling pressure was obtained in specimens prepared by method Inserted, followed by method As-compacted, Transferred and Trimmed. The distinct difference between the swelling pressure could be largely attributed to the effects of residual post-compaction lateral stress. More interestingly, specimens prepared by methods Trimmed and Transferred followed a similar swelling pressure-dry density relationship. From this point of view, method Transferred without causing mass loss, change of bentonite content and possible technological voids effect was recommended in lieu of method Trimmed for specimen preparation in laboratory.

Building the bridge between U(VI) and Ca-bentonite – Influence of concentration, ionic strength, pH, clay composition and competing ions

In the context of a high-level nuclear waste disposal, the retention of U(VI) on non-pre-treated Ca-bentonite as potential technical barrier is studied. The objective of this study is to reveal the retention behaviour of U(VI) under extreme geochemical conditions, such as hyperalkaline pH range as well as high salinity at the same time, and taking into account other relevant parameters. This should lead to a better understanding of necessary safety precautions for avoiding a release of U(VI) in the environment. Batch experiments were conducted to determine the influence of the initial U(VI) concentration, salinity, pH value, clay composition and the presence of other elements (Ca(II), I−, Cs(I), Eu(III)). After the sorption experiments, the remaining U(VI) concentration in solution was determined via mass spectrometry with inductively coupled plasma. U(VI) can be immobilised from 10% to 100% under all investigated conditions. Precipitation plays a role in the U(VI) retention but only at higher concentrations (≥10−5 mol L−1). The retention is reversible especially with decreasing pH (<10.5) as the aquo complex Ca2UO2(CO3)3(aq) is formed. Ca(II) strongly enhances the U(VI) adsorption onto Ca-bentonite in the hyperalkaline pH range, probably due to the formation of Ca(II)-bridges. The best retention could be observed on natural bentonite compared to pure montmorillonite and altered bentonite. From a waste cocktail containing important elements of the repository inventory (Cs(I), Eu(III), U(VI) and iodide), only Eu(III) as homologous element to trivalent actinoids competes with U(VI) for binding sites, especially at low metal concentrations, but also facilitates the precipitation at higher concentrations.

Fractal approach in expansive clay-based materials with special focus on compacted GMZ bentonite in nuclear waste disposal: a systematic review.

Knowledge of the behavior of highly compacted expansive clays, as an engineered barrier, in disposal of high-level nuclear waste (HLW) systems to prevent the pollution due to migration of radionuclide is extremely essential. The prominent properties of globally and widely used bentonites have been extensively studied during past two decades. In China, GaoMiaoZi (GMZ) bentonite is the first choice as a buffer or backfill material for deep geological repositories. This review article presents the recent progresses of knowledge on water retention properties, hydromechanical behavior, and fractal characteristics of GMZ bentonite-based materials, by reviewing 217 internationally published research articles. Firstly, the current literature regarding hydrogeochemical and mechanical characteristics of GMZ bentonite influenced by various saline solutions are critically summarized and reviewed. Then, the role of osmotic suction π alongside the application of surface fractal dimension Ds is presented from the standpoint of fractal theory. Finally, the strength characteristics of GMZ bentonites using fractal approach have been discussed. Furthermore, this study sheds light on gaps, opportunities, and further research for understanding and analyzing the long-term hydromechanical characteristics of the designed backfill material, from the standpoint of surface fractality of bentonites, and implications of sustainable buffer materials in the field of geoenvironmental engineering.

Lead/lead-alloy as a corrosion-resistant outer layer packaging material for high level nuclear waste disposal

Safe disposal of high-level nuclear waste (HLW) is a challenging problem world-wide. Corrosion-resistant waste packages constitute a key component of a multiple barrier system for waste isolation. Metallic copper has been proposed as an outer layer material for a waste package. However, a concern has been raised regarding potential copper corrosion induced by hydrogen sulfide in a reducing disposal environment. We here demonstrate that lead/lead-alloy materials could be an excellent alternative material for waste package outer layers, owing to their corrosion resistance (especially to hydrogen sulfide attack) and radiation-shielding capability. Our long-term (∼800 days) corrosion experiments show that lead could be passivated by a dense layer of sparingly soluble corrosion products, cerussite (PbCO3) and tarnowitzite (Ca,Pb)CO3, in contact with carbonate-bearing groundwaters. If needed, the carbonate concentration in a repository can be conditioned by adding carbonate materials such as calcite or hydromagnesite as a buffer material component to ensure cerussite precipitation. With its low solubility (<10–6 mol•kg−1), cerussite is more favored to form over galena (PbS) in a typical disposal environment. Thus, the issue of sulfide-induced metal corrosion as that happens to copper can potentially be eliminated. Furthermore, using lead/lead alloy will provide excellent radiation shielding for waste package transportation and repository operation.