High-level Nuclear Waste Repository Research Articles

Speciation studies at the Illite - solution interface: Part 2 – Co-sorption of uranyl and phosphate ions

Although it is well known that clays and phosphate ions (hereafter referred to as “P”) can contribute to the long term uptake of uranium (U) in a variety of geochemical systems, work is still needed to document the surface speciation of U(VI) and P sorbed on relevant clay minerals such as illite. Following on from previous work showing strong sorption of phosphate ligands on the surface of a homoionic Na-illite ([1]), we addressed the (competitive/synergistic) mechanisms of (co)sorption of trace levels of uranyl ions (1–10 µM) and phosphate ligands (100 µM) onto this clay, and the identity of the surface species formed, by in situ spectroscopy monitoring of the clay-solution interface along sorption. We also performed complementary batch sorption experiments and electrophoretic mobility (EM) measurements. Macroscopic data indicated a uranyl sorption dependent on pH, aqueous U concentration and clay-to-solution ratio, a signature of P on the U sorption, and a promotion of P sorption with U concentration. Macroscopic and EM data also suggested mostly reversible mechanisms of P and U co-sorption that confer negative charges on the mineral surface and involve several types of uranyl phosphate surface species and/or surface sites present on the clay edges. FTIR interface spectra confirmed that several types of inner sphere uranyl phosphate surface species formed at acidic pH at the illite-solution interface, as a function of U surface coverage and/or reaction time. An inner-sphere uranyl phosphate surface complex, likely with a U-bridging structure, was formed rapidly on high-affinity surface sites existing in limited quantities on the clay edges. Another U-P surface complex (with a possible P-bridging structure) was progressively formed in increasing amounts with U surface coverage and time on low affinity clay edge sites. Finally, a third U-P surface species having an autunite-like structure (probably a U-P polynuclear surface species) was formed on the illite surface at high U concentration (10 µM). The two later species competed with the existence on the illite surface of inner sphere surface complexes and outer-sphere surface complex of phosphate, which were identified in the absence of U. Information on uranyl surface speciation in the presence of phosphate ligands presented here is novel and provides for the first time in situ spectroscopic evidence of synergistic U-P sorption process leading in the formation of several types of uranyl phosphate sorption species on the surface of illite. These results provide a sound basis for better understanding / prediction of U sorption in the environment, including in clayey formations being considered as long term barriers to radionuclide migration in far field of high level radioactive waste repository.

Dissolution of high burn-up spent nuclear fuel at high-pH

Abstract The release of radionuclides from high-burnup spent nuclear fuel (SNF) segments was studied at pH = 13.2 as well as the effect of the presence of calcium and silicon. The aim was to ascertain the dissolution of SNF in solutions corresponding to a high-level nuclear waste repository including concrete in different structural parts. The release of uranium at pH = 13.2 was higher than at pH = 8.4 in bicarbonate medium, while the presence of calcium resulted in a decrease of the uranium concentrations in solutions, assumed to be the consequence of the formation of a secondary solid phase such as Ca2U2O7. Caesium release was found higher at pH = 13.2 as well, but it was not influenced by the presence of Ca and Si at long term. On the other hand, actinide elements (plutonium, neptunium and americium) dissolution decreased at pH = 13.2, probably because of the formation of secondary solid phases. On the contrary, ruthenium and technetium release at pH = 13.2 was found to be much higher than the measured at lower pH, perhaps due to the higher dissolution kinetics of the metallic inclusions at such pH.

Hydride reorientation in fuel cladding under interim storage conditions with low hoop stress

Abstract. The restart of the search for a final repository for high-level nuclear waste in Germany and the resulting so-called white map lead to a reassessment of the safety of interim storage as well. Due to the resulting increase in interim storage periods with a prospected timescale until 2068, among other things, renewed consideration of cladding tube integrity is required for the handling of the fuel assemblies during repackaging into the final storage containers. Besides oxide formation during operation, hydrides can have a negative influence on stability during interim storage and handling later on. The hydrides, which are initially precipitated tangentially in the tube material during operation, can reorient during interim storage as a result of the temperature and load history and precipitate again in a radial direction during cooling under load. In recent years, a number of investigations have already been carried out with respect to this topic, with cladding conditions that are present in high burn-up fuel assemblies with corresponding high hydrogen contents. In these cases, reorientation effects did occur with stresses that can be encountered in interim storage, but the proportions of radial hydrides were very low and were classified as not being harmful. In addition to the tests with cladding tubes containing high hydrogen contents, sporadic tests with low hydrogen contents were also carried out, for example, by Hardie and Shanahan (1975) and later by Chu et al. (2008). These indicated that strong reorientation effects could occur because of the complete dissolution of hydrogen under high temperature and tensile stress. In recent years, Kaufholz et al. (2018) also formulated a hypothesis that describes this effect as being the consequence of free nucleation without the presence of precipitates. This results in the possible formation of extremely pronounced radial hydrides, which can even extend through the tube wall. This is possible because residual hydrides that can hinder free precipitation or represent preferential points for nucleation are absent. The results of the present investigations confirm this behaviour. They show that massive reorientations can already occur at stresses at the lower end of the literature values and that, until now, these reorientations were interpreted as being uncritical. Considering the hysteresis in solubility between terminal solid solubility for dissolution (TSSD) and precipitation (TSSP), it becomes apparent that precipitation at hydrogen contents of 100 ppmw (parts per million by weight) already starts at approx. 538 K and that the hydrides precipitate radially under a stress of 52 MPa and form extremely pronounced hydrides. This results in a need for further research, since the thermal and mechanical loads under which the tests were carried out are well below the permissible limit loads in interim storage and below the values expected so far. This applies in particular to materials and cladding tubes that either absorb less hydrogen during operation or have only low burn-ups. The study provides an overview of the status quo, presents part of the investigations carried out during the project MUDZ (Modellierung und Untersuchung der Degradation von Hüllrohrmaterialien aus Zr-Legierungen durch Hydridbildungs- und Hydridverteilungsprozesse im Hinblick auf die Langzeitzwischenlagerung) in the KEK (Kompetenzerhalt in der Kerntechnik) initiative and the results obtained, and discusses the ramifications of the results.

Communicating scientific uncertainties

Abstract. Communicating a message of scientific uncertainty to a general audience is difficult because laypeople usually lack the conceptual tools to understand uncertainty (e.g., quantification methods). In addition, uncertainties can lead to mistrust and can be misused to prevent or delay radioactive waste management. On the proposed poster, we report the results of an experimental study (N=177) that investigated preferences for and trust in different visualizations of simulation results and tested different modes of communication (verbal, numerical, and visual/graphical). The participants processed three different parts, namely (1) the variation in the form of presentation of a scientific result, (2) the variation in the sender/source of a scientific result in numerical format, and (3) a direct comparison of two representations of simulation results (deterministic and probabilistic). Besides socio-demographics, we used items to check for the potential impact of numeracy and verbal skills, the need for a cognition concept, and questions on science in general, as well as the acceptance of a potential deep-ground repository. The dependent variable in each part was the degree of trust in the scientific validity of the information. For an example result, see Fig. 1. Clear influences of the non-dependent variables are only evident to a limited extent. Women rate their numerical abilities less highly than men do. There is no difference in verbal skills. The latter variable is also not explanatory, while numerical ability is relevant for the choice of the probabilistic variant and the corresponding trust rating in Experiment 3. The most striking result is that the attitude towards a deep geological repository for high-level nuclear waste (HLW; degree of acceptance) is the decisive parameter in all experimental conditions. The oral presentation will focus on another challenge of uncertainty communication, namely unquantifiable model uncertainties. In a repository, these may arise from the heterogeneity of the host rock. One of these potential uncertainties may be an undetected fault parallel in the drift axis in clayey host rocks. At the Konrad site, an anchor failure in an infrastructure drift resulted from such a fault. To deal with this kind of uncertainty, a certain kind of flexibility is needed to react to such events. The talk should initiate a discussion on the importance of communicating these potential uncertainties and on the impact of these uncertainties on the repository program.

Stainless steel corrosion under anoxic, highly saline and elevated temperature conditions

Abstract. Several countries consider hosting a deep geological repository for high-level nuclear waste (HLW) in salt rock. In the unexpected case of solution access to the emplacement caverns during the long-term evolution of such a repository, metallic containers will be exposed to highly saline brines. In this study, corrosion experiments under conditions expected to be representative of HLW disposal in salt rock have been performed with stainless steel, a material typically used to construct containers containing vitrified HLW. Experiments were performed in closed vessels under anoxic and highly saline conditions at 90 ∘C. Polished stainless steel coupons were suspended in 5 mol L−1 NaCl or 3.4 mol L−1 MgCl2 at their natural pH. At the end of the experiments (up to 294 d), vessels were cooled down to room temperature, pH and Eh were measured in situ, the ultra-centrifuged contacting brines were analyzed by (HR)ICP-MS (high-resolution inductively coupled plasma mass spectrometry) and the formed corrosion products were identified. In all corrosion experiments, pHm and Eh hardly changed with exposure time, and the nature of the salt had no significant effect. The overall corrosion rates were low, and the quantities of dissolved metal ions revealed the formation of only sparingly soluble corrosion products. The formation of a passivation layer mostly made of Cr(III) (hydr)oxide was evidenced, but no localized surface attack could be detected. The coupon corroded in 5 mol L−1 NaCl was embedded in resin and cross-cut for further analyses, using synchrotron-based techniques with a small beam footprint (X-Ray fluorescence (μXRF), X-Ray diffraction (μXRD), and X-ray absorption near-edge structure (μXANES)). Element distribution maps revealed the presence of a very thin (<1 µm) layer of corrosion products, with regions enriched either in Cr or in Fe and Ni. X-ray diffractograms identified the presence of spinel-type compounds (Fe3O4 and NiFe2O4) and Cr2O3, which was corroborated by recording μXANES at the Cr, Fe and Ni K-absorption edges. The additional presence of layered double hydroxides in addition to NiO and Ni(OH)2 was also evidenced. Overall, results point to a corrosion layer having a duplex structure, with an inner layer mostly of chromium (hydr)oxides and an outer layer made of Fe- and Ni-based spinel compounds admixed with nickel (hydr)oxides.

SpannEnD – a 3D geomechanical model of Germany for the prediction of the recent crustal stress state

Abstract. The recent crustal stress state is a crucial parameter in the search for a high-level nuclear waste repository. It is relevant not only for site selection and short-term stability during operation but also for predicting long-term safety. However, the level of knowledge in this regard is still limited in Germany. It is based on stress orientations and stress regime information from the World Stress Map (WSM) project and pointwise stress magnitude records. We present the results of a 3D geomechanical numerical model that improves the state of knowledge by providing a continuum-mechanics-based prediction of the recent crustal stress field in Germany. The model extends over an area of 1000×1250 km2, covering Germany and nearby regions. It contains 22 geological units, each parameterized with individual properties (density, Young's modulus and Poisson's ratio). We assume linear elasticity, and the finite element method is used to solve the equilibrium of forces. The model is calibrated with horizontal stress magnitude records and validated with additional data, for example, vertical stress magnitudes or data of the WSM project. Our model results are in a good agreement with a mean orientation of the maximum horizontal stress (SHmax) derived from the WSM, as indicated by a mean of the absolute differences of 12∘. Furthermore, the model results lie entirely within the standard deviation of the derived orientation of SHmax. Mean values of the absolute stress differences between calibration data and model results of 4.6 MPa for the minimum horizontal stresses (SHmin) and 6.4 MPa for the SHmax also show a good agreement. The model results can be used, for example, for the calculation of fracture potential, for slip tendency analyses or as boundary conditions for smaller local models. For smaller model volumes, where little or no stress information is available, it provides synthetic stress magnitude data that can be used for calibration. As such, the model provides a valuable starting point for the detailed assessment of stability at yet-to-be-determined siting regions for a high-level nuclear waste repository.

Numerical simulation of long-term performance of deep geological repository placement rooms in crystalline and sedimentary rocks

Evaluation of the stability of placement rooms and damages in the geosphere surrounding a deep geological repository (DGR) for high-level nuclear waste are essential components for assessing the overall repository performance. Assessments of the performance of placement rooms for the duration of the regulatory period (typically 1 million years) usually involves simulations using numerical codes capable of modeling coupled thermo-hydro-mechanical (THM) processes induced by different natural and repository-induced perturbations over different time and length scales. The numerical constitutive model used to represent the host medium should be capable of capturing highly nonlinear responses to stress changes, including development of damage and fracturing. In this study, a novel approach for simulation of damage evolution and fracturing around the repository rooms, based on the bonded block model (BBM) devised for analyzing coupled THM processes, is used to perform a study of the long-term geomechanical performance of the Nuclear Waste Management Organization hypothetical DGR configurations in sedimentary and crystalline geological settings. The perturbation effects of different loading conditions, including in-situ stress changes by room excavation, heat generated from the used fuel canisters, and glaciation were analyzed. Each THM analysis was conducted using a 3D continuum numerical code to generate evolving temperature and pore pressure fields in the rock mass. The temperature and pore pressure data were then imported in the 2D BBM model for simulating evolution of damage around the placement rooms. Sensitivity analysis was conducted to investigate the effect of different poro-mechanical parameters, rock mass permeabilities, and different approaches for calibrating the BBM, on predictions of room stability and extent of the excavation damage zone around the placement room. The modeling results indicate that the confinement provided by the bentonite backfill inside the room clearly prevents any unravelling but also limits the extent of the damage to less than 1.6 m into the room walls in any of the analyzed cases considering different loading conditions and uncertainty of the material properties.

TransPyREnd: a code for modelling the transport of radionuclides on geological timescales

Abstract. The German site selection procedure for a high-level nuclear waste repository is entering a stage in which preliminary safety assessments have to be conducted and the release of radionuclides has to be estimated for a large number of potential sites. Here, we present TransPyREnd, a 1D finite-differences code for modeling the transport of radionuclides in the subsurface at geological timescales. The code simulates the processes advection, diffusion, equilibrium sorption, decay of radionuclides, and the build-up of daughter nuclides. We summarize the modeled physical processes, their mathematical description and our numerical approach to solve the governing equations. Finally, two simple tests are shown, one considering diffusion, sorption, and radioactive decay, the other involving diffusion and a radioactive decay chain. In both tests, the code shows good agreement with the reference solutions. Caveats of the model and future additions are discussed.

The final countdown to site selection for Canada’s nuclear waste geologic repository

ABSTRACTSelecting a site to build a deep geologic repository for high-level nuclear waste is a process fraught with social tension and few countries have managed to do so successfully. Canada is poised to be next, with a site selection decision due in 2024. Two potential sites are under consideration, one in northern Ontario and one in southern Ontario, each with different geologic settings and different social and economic landscapes. We review the consent-based process that has led to this point, the complexities involved in expressing consent in Canada, and some of the hurdles that remain.

A systematic review of bentonite/concrete interaction system in HLW disposal repositories: Theoretical, experimental and geochemical modelling analysis

The buffer/backfill material of bentonite acts the most important role for the safety of high-level nuclear waste repository. After the closure of repository, the concrete as another barrier in geological repositories will degrade and release alkaline components into the site water, which has significant impact on the compositions of bentonite and threatens the long-term safety of repository. Many studies have been carried out to detect the mineralogy alteration of bentonite during alkaline disturbing including laboratory tests, in situ experiments and geochemical modelling. It was found that montmorillonite dissolves and forms low-swelling minerals such as mixed layer clays, non-swelling zeolite and calcium silicate hydrate minerals, etc. The researches that have been done are extensive and disorganized, however, the results are mostly qualitative and far from the engineering requirements.

Experimental Determination of Polycrystalline Salt Rock Thermal Conductivity, Diffusivity and Specific Heat From 20 to 240°C

Salt rock is recognized as one of the most suitable parent rocks for geological disposal of high level radioactive waste due to its low permeability, good ductility, good thermal conductivity and damage self-healing properties. The thermal conductivity of salt rock directly affects the temperature of disposal storage and surrounding rock, high temperature will lead to a series of problems such as nuclear waste storage tank rupture, mechanical and permeability reduction of surrounding rock. In this paper, the thermal conductivity, specific heat and thermal diffusion coefficient of NaCl single crystal made by crystallization method in the laboratory and polycrystalline salt rock from Khewra salt mine were measured in the range of 22–240°C by the transient plane source method. The results showed that the thermal conductivity and thermal diffusivity of salt rock decrease gradually with the increase of temperature, while the specific heat capacity increases with the increase of temperature. The thermal conductivity of salt rock is slightly lower than that of single-crystal NaCl. The reason for this phenomenon may be that there are a few pores in salt rock. Based on the experimental data, the models of thermal conductivity, thermal diffusivity and specific heat of salt rock with temperature were established. The results can provide a reference for the construction of underground salt rock high-level radioactive waste disposal repository. Based on the thermal conductivity model of polycrystalline salt rock established in this paper, the temperature field evolution during the operation of the underground salt rock high-level nuclear waste repository in 1,000 years was studied. It was found that the temperature of the glass solidified of high-level radioactive waste reached the highest (177.6°C) and then dropped rapidly. The decay heat radiation influence radius of the nuclear waste reaches its maximum in about 50 years of operation of the repository and then gradually decreases.

Experimental Study on Physical and Mechanical Properties of Beishan Granite after Cyclic Heating and Cooling with Water

Abstract Granite characterized by low permeability and high strength is commonly seen in underground construction. This paper is aiming at studying the influence of cyclic heating and water cooling on physicomechanical properties of Beishan granite, which is chosen as a potential host surrounding rock of a high level nuclear waste (HLW) repository. In this study, physical and mechanical properties of Beishan granite after repeated heating-cooling cycles are described. After five heating-cooling cycles with temperature up to 300°C, the diameter and length of specimens increase by 0.090% and 0.052%, respectively. And the mass of specimen reduces by 0.32 g. In the following conventional triaxial compression (CTC) tests, the peak strength shows a slightly increase, but the dilatancy threshold decreases. And both peak strain and dilatancy threshold strain exhibit increase, which is responsible for the decrease of Young’s modulus. During the test, acoustic emission (AE) is rarely monitored in the early loading and starts to generate systematically when the axial stress is beyond dilatancy threshold. On the other hand, in the hydraulic (HM) coupled experiments, both peak strength and dilatancy threshold show decrease. However, the ratio between dilatancy threshold and peak strength shows slightly increase. The same phenomenon can be observed for the peak strength and dilatancy threshold. The permeability shows rarely change compared with that of the specimen without experienced cyclic heating-cooling. AE is significantly monitored in the initial loading, which is seem more than that generated from dilatancy threshold to peak strength. The obtained results are useful for the construction of HLW repository.

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.

Aggregated germanium saponite: Removal and retention of polymeric thorium and uranium complexes

The present study concerns the removal of thorium and uranium from aqueous solution, and particularly the effect of clay particle size and morphology on the adsorption of polynuclear actinide complexes by synthetic saponite nanoscale powders. Saponites were synthesized from a reactant mixture containing aluminum nitrate, magnesium nitrate, and either sodium silicate or sodium germanate. The saponite powders were analyzed by TEM, XRD, SAXS, N2-adsorption-desorption isotherms and FT-IR. Ge-saponite contained randomly stacked platelets ≈ 100 nm long and 80 nm wide, while Si-saponite showed roughly spherical agglomerates ≈ 60–80 nm in size. Small-angle X-ray scattering measurements indicated a hierarchical pore network among the plates and their agglomerates. Thorium polymers composed of fractal aggregates of branched filaments were characterized by synchrotron radiation-SAXS. The distribution coefficient (Kd mL/g) values for thorium (20 μg/mL) at pH 4 were: Ge-saponite (105) ≈ Si-saponite (105) > raw saponite (104) ≈ raw saponite (washed) (104). The Kd values for uranium (20 μg/mL) at pH 5.5 were: Ge-saponite (103) ≈ raw saponite (washed) (103) ≈ Si-saponite (103) ≈ raw saponite (103). The difference in uranyl adsorption performance observed between the two nanoparticle forms of saponite was attributed to a larger pore width for Ge-saponite (3.5 nm) compared with Si-saponite (0.8 nm). The results have implications for mineral ore tailings containing thorium(IV) and, by analogy, Pu(IV) in a clay buffer environment within a geological repository for high-level nuclear waste.

Alternative Disposal Options for High-Level Radioactive Waste

Abstract. Are alternatives to the disposal of high-level radioactive waste in a geology repository conceivable? We present the results of the first phase of a research project on the state of the art in science and technology for alternative disposal options. The project is financed by the Federal Office for the Safety of Nuclear Waste Management. Most recently, in 2015, the German Commission on the Storage of High-Level Radioactive Waste (Endlagerkommission) evaluated possible disposal technologies and classified them as either promising, conceivable, or to be pursued further. Only final disposal in a geological repository was considered promising. Conceivable, but not immediately available or not advantageous, were storage in deep boreholes (DBs), long-term interim storage (LTIS), and partitioning and transmutation (P&T). All other alternative disposal options by burial, dilution, or removal from the planet were determined not to be worth pursuing. The Disposal Commission did conclude that none of the three conceivable methods (DBs, LTIS, P&T) would result in earlier disposal of high-level radioactive waste than the preferred final disposal in a mine. However, it recommended continued tracking and regular monitoring of the future development of alternative disposal options, e.g., disposal in deep boreholes. Finally, in 2017, with the amended Site Selection Act, the federal government specified disposal in a repository mine with the option of retrieval during operation or recovery for 500 years after closure. In a learning site selection process, the Federal Office for the Safety of Nuclear Waste Disposal (BASE) reviews the proposals of the project managing company, the Federal Company for Radioactive Waste Disposal (BGE), and prepares a reasoned recommendation to the federal government for a site with the best possible safety. Part of the reasoned recommendation is, among other things, a discussion of alternative disposal options to final disposal in deep geological formations. In the presentation, we report on the status of international research on alternative disposal options, discuss advantages and disadvantages of the technologies, and evaluate the potential of the technologies for the disposal of high-level radioactive waste in Germany. The LTIS is designed as dry storage in a building to be constructed above ground or near the surface and is expected to last for a period of several hundred years. With LTIS it would be possible to gain time for the development of a suitable final disposal option; however, this also postpones the disposal issue indefinitely into the future with undetermined methods. DB storage would involve sinking the storage containers into boreholes with depths of up to 5000 m. This could reduce the expense and be particularly advantageous for smaller inventories, although the potential for the use of engineered barriers would be limited and retrievability precluded according to the current state of the art in science and technology. P&T is primarily intended to separate long-lived transuranic elements from high-level radioactive waste and then convert them to short-lived fission products by neutron irradiation in reactors. The main goal is to reduce the necessary containment times in the repository by changing the inventory, but the effort to treat the waste would be significant and a repository for high-level nuclear waste is still needed. More exotic ideas for alternative disposal include deep geological injection of liquid waste, waste forms that melt themselves into rock, storage inside the ocean floor or subduction zones, shipment to space, burial in ice sheets, or dilution in the atmosphere and oceans. None of these exotic options is currently being actively pursued.

Geoscientific Characterization and Interpretation (Geosynthesis) within the Preliminary Safety Assessment in the German Site-Selection Procedure for a High-level Nuclear Waste Repository

Abstract. After implementation of the Repository Site Selection Act (StandAG) in 2017, the Federal Company for Radioactive Waste Disposal (BGE), as the German waste management organization, started the site selection procedure for a nuclear repository for high-level radioactive waste in Germany. On the way towards the repository site with the best possible safety, the site selection procedure is required to be a participatory, transparent, learning and self-questioning process based on scientific expertise. With the Subareas Interim Report published in 2020, first results were presented outlining subareas with favourable geological conditions in preparation for defining the siting regions for surface exploration. Currently, one of the main tasks in the site selection procedure is to establish a detailed geoscientific synthesis (Geosynthesis) for each subarea. The Geosynthesis contains all geological information for the characterization of each subarea and hence serves as the foundation for the subsequent analysis within the representative preliminary safety assessments (rvSU) and the geoscientific consideration criteria. Based on this information, all areas within the subareas will be evaluated to find the siting regions for surface exploration. The Geosynthesis includes a description of the regional geology focusing on the host rock, the overburden and relevant geological processes that may affect the potential nuclear waste repository in the next 1 million years. The data for the Geosynthesis are mostly compiled from state authorities and include 3-D geologic models, regional maps and cross-sections, bore hole data (e.g. geophysical logs) and seismic data. Furthermore, it is necessary to digitize, process, interpret and evaluate the aforementioned data using the available knowledge from the scientific literature in the context of the site selection procedure.

An Open-Access Stress Magnitude Database for Germany

Abstract. The stress field in the Earth's crust plays a central role in the site-selection process for a deep geological repository for high-level nuclear waste. Site selection and construction planning must take into account several factors that are influenced by the stress state. These include the excavation damage zone, the hydraulic permeability of the host rock, the self-sealing capacity, the effects of seismic events and the possible reactivation of faults as migration pathways for fluids and radionuclides. Likewise, the initial stress state is of central importance for the long-term studies to prove site safety over 1 Ma. To obtain a continuous description of the current 3D stress state, 3D geomechanical numerical models are used. These models have to be calibrated with data on stress magnitudes to obtain robust predictions. One of the central goals of the SpannEnD project (Spannungsmodell Endlagerung Deutschland, http://www.spannend-projekt.de, last access: 31 October 2021) was to build the first comprehensive and publicly accessible stress magnitude database for Germany, including a quality ranking of the data compiled from different methods. This database is the logical extension of the database of the World Stress Map project, in which so far only information on stress orientations and the stress regime has been compiled systematically. We present this first compilation of stress magnitude data published and made available by Morawietz et al. (2020). The stress data density is generally low and heterogeneous, so that a model calibration at the scale of a site model is not possible. Therefore, the main objective of the SpannEnD project is to develop a 3D geomechanical numerical model for the whole of Germany. The resulting 3D stress field will provide the basis for regional and local models in a later phase of the site selection process. Details on this are presented in three complementary contributions in this symposium by Reiter et al., Röckel et al. and Ahlers et al. The new Geology Data Act (Geologie-Datengesetz) now allows access to considerably more data, which will be incorporated into an update of the database after assessment according to the defined quality criteria. This database extension will improve the reliability of the predictions of the geomechanical models on different spatial scales.

Non-isothermal water retention curve model for adsorption

Non-isothermal water retention curve (WRC) in high suction range is a fundamental and interpretative tool to describe the behavior of bentonite and is crucial to ensure the safety and effective operation of high-level nuclear waste repository. In this study, an adsorption equation at given water content is proposed according to Kelvin equation, Clausius-Clapeyron equation and the adsorption enthalpy equation. Then, the proposed adsorption equation is incorporated into van Genuchten-Mualem model to establish a non-isothermal WRC model for adsorption. Four series of non-isothermal WRC data in a temperature range (293∼353 K) and high suction range (>1 MPa) from four different soils are employed to illustrate the application and performances of the proposed model. The proposed WRC model provides good agreement between the calibrated/predicted and measured non-isothermal WRCs.

Trust, mistrust and distrust as blind spots of Social Licence to Operate: illustration via three forerunner countries in nuclear waste management

The notion of social licence to operate (SLO) has become a widely applied concept for companies in mining and resource extraction industries to manage their social and community relations, in the face of local criticism and opposition. SLO literature and practice have highlighted earning the trust of the local community as a key requirement for an SLO. This article addresses three weaknesses in how the current SLO literature addresses trust. The arguments are illustrated via examples from nuclear waste management in Finland, France and Sweden – three forerunners in implementing high-level nuclear waste repository projects. Nuclear waste management constitutes a relevant case for analysis, as an industry that faces significant risk-related challenges of local acceptance, ethics, economics, and democratic debate. Focussing on the oft-used SLO framework of Boutilier and Thomson, with its emphasis on interactional and institutionalised trust between the company and the local community, we address three gaps in the SLO literature: 1) insufficient conceptualisation of trust, in particular the dynamics between different dimensions of trust, mistrust and distrust; 2) lack of attention to the potential virtues of mistrust and distrust; and 3) the downsides of taking the institutionalisation of trust as the ultimate criterion of a strong SLO, especially in contexts entailing significant asymmetries of power. The article concludes by suggesting ways of alleviating the identified weaknesses, via greater recognition of the multidimensionality of trust, mistrust and distrust, the virtues of mistrustful civic vigilance, and greater attention to trust dimensions that lie beyond the community-company relations.

Theoretical and Numerical Study on Electrical Resistivity Measurement of Cylindrical Rock Core Samples Using Perimeter Electrodes

The estimation of hydraulic and mechanical properties of bedrock is important for the evaluation of energy-related structures, including high-level nuclear waste repositories, hydraulic fracturing wells, and gas-hydrate production wells. The hydraulic conductivity and stress–strain curves of rocks are conventionally measured through laboratory tests on cylindrical samples. Both ASTM standards for hydraulic conductivity and compressive strength involve the use of the planar bases of a cylindrical sample. Hence, an alternative test method is required for the simultaneous measurement of hydraulic conductivity and stress–strain curves. This study proposes a novel electrical resistivity estimation method using two perimeter electrodes for the estimation of hydraulic properties. The theoretical background for the perimeter electrode setup is derived and the COMSOL MultiPhysics® finite element numerical simulation tool is employed to verify the derived theoretical equation. The accuracy of the numerical simulation tool is first validated by simulating the ASTM standard testing method for electrical resistivity. The electrical resistance values derived from the theoretical equation and numerical simulation are compared for different electrical resistivity and electrode radius. The assumed equidistant, circular equipotential surface results in a theoretical lower bound for the measured electrical resistance in the cylindrical specimen. The introduction of a phenomenological distortion factor to correct for the theoretical equipotential surface results in a good fit with the numerical simulation results. The effects of electrode length and equivalent strap electrodes were investigated to assess the applicability of the suggested method for laboratory testing. Consequently, this study presents an effective alternative theoretical assessment method for the lower bound electrical resistivity of cylindrical rock core samples under confining conditions when the installation of base electrodes is infeasible.