The safety concept of a repository in rock salt is based on a multi-barrier system in that barrier’s sealing effects are time dependent. In short-term of the post-closure phase, the safe containment of radionuclides is provided by the waste canisters and geotechnical sealing elements (e.g., drift and shaft seals). In long-term, the sealing effect is provided by the geological barrier together with the crushed salt backfill. The sealing effectiveness of crushed salt evolves with ongoing compaction and therefore reduction of porosity/permeability. For the long-term safety assessment, the period in which crushed salt reaches barrier properties is a crucial information.

Numerical simulations must be used to predict the long-term compaction behaviour of crushed salt, thus it needs confidence in the results. Confidence can be built by extensive testing, verification and validation of individual simulation codes and the underlying constitutive models and by comparing constitutive models with different approaches/backgrounds.

Within the MEASURES project, a systematic approach was chosen to build trust in numerical simulations. A calculation task is defined including the analysis and calibration of constitutive models against systematic experimental observations and the subsequent calculation of a generic backfilled drift providing the opportunity to directly compare the model performance and strikingly visualize the differences. The aim is to achieve a convergence of the simulation results by systematically improving the material models and adapting the parameters to the growing experimental observations.

The process of calibration and model comparison is an iterative process and performed several times and the continuous progress is shown by the simulation of a generic backfilled drift. Each time a calibration process is finished, the generic backfilled drift is modelled and the evolution of porosity (identified as most important process variable) over time is evaluated.

The importance of employing different simulation codes/constitutive models and comparing their results is based on confidence building in the numerical predictions, quality assurance of the models and giving robustness to the results. Further, there is no complete consensus on formulating constitutive models. In MEASURES, the models are divided into phenomenological based and microstructural based formulations (Table 1) (Friedenberg et al., 2024). Uncertainties are quantified by using different model formulations due to the possibility to interpret the diversity and complexity of many physical processes involved and due to a general comparison of different methods.

Table 1. Overview of constitutive models and simulation codes

 Acknowledgements

Thanks go to the MEASURES family for the constant support.

The project partner GRS, BGE-TEC, IfG and TUC acknowledge the project funding received by the German Federal Ministry for the Environment, Nature Conservation, Nuclear Safety and Consumer Protection (BMUV), represented by the Project Management Agency Karlsruhe (PTKA) (FKZ 02 E 12214 A-D).

Sandia National Laboratories is a multimission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-NA0003525.

References

Friedenberg, L., et al. (2024). Kompass-II: Compaction of crushed salt for safe containment - phase 2 (GRS-751). 

How to cite: Friedenberg, L., Gartzke, A.-K., Lerch, C., Lerche, S., Li, L., Liu, W., Rahmig, M., Reedlunn, B., and Saruulbayar, N.: Confidence building in the prediction of crushed salt as long-term barrier, Third interdisciplinary research symposium on the safety of nuclear disposal practices, Berlin, Germany, 17–19 Sep 2025, safeND2025-25, https://doi.org/10.5194/safend2025-25, 2025.

A widely-proposed approach to dispose low- and intermediate-level radioactive waste is to store it in a deep underground repository with a multiple barrier system. The outer barrier consists of the combination of the host rock along with appropriate backfill materials. The inner barriers include steel drums and emplacement containers. Cement has been found to support a high pH environment that is favourable for radionuclide retention, as well as to suppress microbial activity and slow down metal corrosion. However, different gases, such as hydrogen, methane and carbon dioxide, can be produced during the process, with the latter being absorbed by a cement carbonation reaction. Degradation of organic waste and metal corrosion will consume water while the produced gasses lead to a local pressure build-up which in turn may leads to reduced water supply suppressing further gas generation. Understanding this feedback system and the geochemical evolution of the barriers, and assessing the maximum pressure build-up, is critical to the performance assessment of the repository.

In this case study, we use the coupled reactive transport model of component based two-phase flow in the OpenGeoSys framework to simulate a 2-dimensional cross-section of a disposal gallery, which is following the Swiss disposal strategy in a low permeable clay-type host rock. In this strategy, several concrete containers filled with metal waste or steel drums of organic waste are stacked into a gallery, which is then mostly backfilled with low capillarity, high porosity mortar with low initial water saturation. In this concept, the mortar should buffer gas production, but due to low capillarity, the full re-saturation of the gallery will take thousands of years. We have implemented a geochemical model that treats the degradation of different cement materials with a lookup-table approach. These tables store pre-calculated changes in porosity, the consumption and release of water and gases, and change in cement pore water pH.

In this presentation, we show simulation results covering the geochemical evolution of a gallery cross section over several thousand years. We quantify the local gas generation rates within the waste packages and show how gas will be distributed within the gallery cross section and how gas components will dissolve and dissipate within the host rock. We have performed simulations with different clay rock permeabilities and initial liquid saturation of the mortar and the organic waste packages to investigate maximum pressure build-up. We also show how gas generation rates change when metal waste containers are stored either at the bottom or at the top of the gallery. Our results show that simulations of the whole repository could be improved by using local gas generation rates instead of using estimates of gas generation rates based only on waste inventories and chemical reaction kinetics.

How to cite: Vehling, F.: Two-phase reactive transport modelling of gas production and pressure build up over a gallery cross section in a low-level radioactive waste repository, Third interdisciplinary research symposium on the safety of nuclear disposal practices, Berlin, Germany, 17–19 Sep 2025, safeND2025-31, https://doi.org/10.5194/safend2025-31, 2025.

Given the legal requirement in Germany for documenting uncertainty impacts on safety analyses (§11 EndlSiUntV), a systematic investigation of uncertainties in simulation results is imperative. In several projects, including ANSICHT-II [1], MeQUR [2], and ThermoBase we have investigated the forward propagation of input parameter uncertainties through numerical models. Since modeling studies on long-term safety itself, such as on the host rock integrity of repository systems, represent computationally challenging problems, methods for quantifying the uncertainties have been enhanced and adapted to the particular requirements.

This study delineates a two-step process for uncertainty quantification, commencing with variance-based sensitivity analyses to ascertain the significance of individual input parameters on the integrity of the containment-providing rock zone (CRZ). For this purpose, all numerical input parameters are first considered uncertain. Bandwidths for each parameter are estimated using literature values and expert knowledge. Sobol indices are then determined, and key parameters whose uncertainty significantly influences results are identified. Subsequently, these parameters are employed to define a reduced stochastic state space, which is explored through techniques such as Monte Carlo sampling and stochastic collocation. The reduced state space enables comprehensive stochastic evaluations on full-scale models with only a minimal reduction of the total variance.
Within the ANSICHT-II project, criteria in alignment with legal safety requirements were developed to evaluate CRZ integrity in clay rock. These criteria establish a functional relationship between uncertain input parameters and simulation output. The two-staged process is conducted with the ANSICHT NORD model as an example and the integrity of the CRZ described by criteria functions as quantities of interest. 

Specialized software crafted to meet the computational challenges associated with uncertainty quantification in numerical integrity analyses of repository systems were developed. Notably, this is the OpenGeoSys Uncertainty Quantification framework (OpenGeoSysUncertaintyQuantification.jl [3]) in the Julia language. This enables the stochastic postprocessing of such analyses without prior data reduction. Finally, an interactive dashboard designed to provide users with intuitive access to the uncertainty quantification results, thereby enhancing the transfer of knowledge regarding safety-relevant processes in repositories and their inherent uncertainties, is presented.

[1] J. Maßmann, J. et al. (2022). Methode und Berechnungen zur Integritätsanalyse der geologischen Barriere für ein generisches Endlagersystem im Tongestein. Projekt ANSICHT-II. Ergebnisbericht. Bundesanstalt für Geowissenschaften und Rohstoffe (BGR).

[2] Kurgyis, Kata, et al. "Uncertainties and robustness with regard to the safety of a repository for high-level radioactive waste: introduction of a research initiative." Environmental Earth Sciences 83.2 (2024): 82.

[3] Bittens, M. (2024). OpenGeoSysUncertaintyQuantification.jl: a Julia library implementing an uncertainty quantification toolbox for OpenGeoSys. Journal of Open Source Software, 9(98), 6725.

How to cite: Thiedau, J., Bittens, M., and Maßmann, J.: Computational Integrity Analysis: Approaches for Uncertainty Quantification and Visualization, Third interdisciplinary research symposium on the safety of nuclear disposal practices, Berlin, Germany, 17–19 Sep 2025, safeND2025-140, https://doi.org/10.5194/safend2025-140, 2025.

Stochastic discrete fracture network (DFN) models are commonly employed to represent fractured media, improving the reliability of hydro-mechanical coupled models. DFNs have been established as an essential asset to ensure post-closure safety of nuclear waste deposits in crystalline host rocks [1]-[5].
However, a comprehensive understanding of fracture patterns and statistical properties often requires more than borehole investigations. Due to their quasi one-dimensional nature and limited availability, well-log fracture data present an incomplete, biased picture, which requires assumptions on spatial distribution and clustering of fractures, as well as geometrical properties such as fracture length and aperture distribution to include into a stochastic framework for the modeling process [4]. Furthermore, they do not provide information on structural features, such as fracture length, clustering, and spatial distribution.
To address these limitations, surface data is commonly used as supplementary information to calibrate DFN models [3]. Yet, while two-dimensional surface data offers accessible fracture information, it remains underutilized compared to subsurface data. This is partly due to the influence of weathering and glaciation, as well as surficial variations in stress and strain, which can distort fracture statistics derived from surface observations [2][5].
Here, we explore how fracture properties and patterns evolve with depth through both analytical and numerical approaches. We systematically assess a comprehensive dataset covering remote sensing and surface data, borehole data up to a few hundred meters, and data compiled from the Grimsel Test Site (GTS) plus adjacent tunnels of the Kraftwerke Oberhasli AG (KWO) in Switzerland.
By statistically comparing these datasets and correlating them with known stress conditions and structural models of the region, we aim to mathematically explain the variation of fracture statistics with depth. We aim to clarify when and under what conditions surface data can be reliably used, while also providing a methodological framework for its effective integration into DFN modeling. Our findings may improve the value of surface data, provided there is a robust understanding of the regional structural inventory.

[1] Lavoine Etienne, Davy Philippe, Darcel Caroline, Munier Raymond, A Discrete Fracture Network Model With Stress-Driven Nucleation: Impact on Clustering, Connectivity, and Topology, Frontiers in Physics 8 (2020).

[2] Lee Hartley, Simon Libby, Tomas Bym, James Carty, Mark Cottrell, Kyle Mosley, Baseline Forsmark – A discrete fracture network (DFN) model applying grown fractures and hydromechanical (HM) coupling, SKB Report R-23-01 (2024).

[3] Qinghua Lei, John-Paul Latham, Chin-Fu Tsang, The use of discrete fracture networks for modeling coupled geomechanical and hydrological behavior of fractured rocks, Computers and Geotechnics 85 (2017).

[4] Weiwei Zhu, Xupeng He, Ryan Kurniawan Santoso, Gang Lei, Tad Patzek, Moran Wang, Enhancing Fracture Network Characterization: A Data-Driven, Outcrop-Based Analysis, Earth Space Sci. Open Arch. 35 (2021).

[5] Kevin Bisdom, Bertrand Gauthier, Giovanni Bertotti, Nico Hardebol, Calibrating discrete fracture-network models with a carbonate three-dimensional outcrop fracture network: Implications for naturally fractured reservoir modeling, AAPG Bulletin 98 (2014).
   

How to cite: Perkams, S., Achtziger-Zupancic, P., Amann, F., and Dietl, C.: Comparison of surface and subsurface data for the construction of a comprehensive DFN model, Third interdisciplinary research symposium on the safety of nuclear disposal practices, Berlin, Germany, 17–19 Sep 2025, safeND2025-143, https://doi.org/10.5194/safend2025-143, 2025.

Reactive transport models taking into account water-rock interactions are used to simulate the migration behaviour of radionuclides at potential disposal sites for highly radioactive waste. Previous studies on the Opalinus Clay at Mont Terri (Switzerland) show that hydrogeochemical differences between the host rock and adjacent aquifers caused the development of gradients in pore water geochemistry across the entire system [1,2]. It was found that some of the inherent heterogeneities in this setting, e.g. in ionic strength, significantly affect uranium sorption and thus migration distances [3]. However, the hydrogeochemical conditions in the surrounding aquifers of a host rock are subject to spatial and temporal uncertainties. To evaluate the sensitivity of uranium migration distances in this respect, we performed scenario-based simulations for a period of one million years using the geochemical code PHREEQC.

We quantified the effects of potential brine and seawater intrusion, freshwater enrichment, and acidification, for example, by significantly altering the ionic strength (from 0 to 5 mol/L) and pH (from 3 to 11) in the aquifers at the model boundaries. Results were compared to a reference case based on current conditions at Mont Terri [3]. Simulated uranium migration distances changed by a few metres among the scenarios tested. The primary control is the concentration of the mobile, aqueous ternary uranyl complexes. The complex formation is governed by alkalinity and availability of Ca and Mg, whereby uranium sorption is decreased or increased [4]. Our findings demonstrate that mineral reactions in the Opalinus Clay system at Mont Terri largely buffer hydrogeochemical perturbations in the bounding aquifers within a period of one million years after repository closure. This reduces the sensitivity of the uranium migration distances. Future uncertainty analyses of this kind should include other radionuclides such as neptunium, which may react differently to hydrogeochemical impacts.

[1] Mazurek, M., Alt-Epping, P., Bath, A., Gimmi, T., Waber, H. N., Buschaert, S., De Cannière, P., De Craen, M., Gautschi, A., Savoye, S., Vinsot, A., Wemaere, I. and Wouters, L. (2011): Natural tracer profiles across argillaceous formations. Applied Geochemistry 26 (7), 1035-1064. DOI: 10.1016/j.apgeochem.2011.03.124

[2] Pearson, F. J., Arcos, D., Bath, A., Boisson, J.-Y., Fernández, A. M., Gäbler, H. E., Gaucher, E. C., Gautschi, A., Griffault, L., Hernán, P. and Waber, H. N. (2003): Mont Terri Project – Geochemistry of water in the Opalinus Clay formation at the Mont Terri rock laboratory. Reports of the FOWG, Geology Series no. 5. Bern, Switzerland: Federal Office for Water and Geology (FOWG)

[3] Hennig, T. and Kühn, M. (2021): Potential uranium migration within the geochemical gradient of the Opalinus Clay system at the Mont Terri. Minerals 11 (10), 1087. DOI: 10.3390/min11101087

[4] Hennig, T., Stockmann, M. and Kühn, M. (2020): Simulation of diffusive uranium transport and sorption processes in the Opalinus Clay. Applied Geochemistry 123, 104777. DOI: 10.1016/j.apgeochem.2020.104777

How to cite: Schöne, T., Hennig, T., and Kühn, M.: Hydrogeochemical impacts on uranium migration in the Opalinus Clay at Mont Terri, Third interdisciplinary research symposium on the safety of nuclear disposal practices, Berlin, Germany, 17–19 Sep 2025, safeND2025-151, https://doi.org/10.5194/safend2025-151, 2025.

Assessment and Communication of Uncertainties in the Site Selection Procedure (BewUSt)

Gerd Frieling¹, Thomas Beuth¹, Esther Kähler¹

¹Federal Office for the Safety of Nuclear Waste Management (BASE), Wegelystr. 8, 10623 Berlin, Germany

Correspondence to: G. Frieling (gerd.frieling@base.bund.de)

Abstract. The German site selection procedure (StandAV) examines whether a safe containment of high-level radioactive waste can be expected in the respective repository concept for one million years [1]. This requirement is assessed in particular with the help of preliminary safety investigations (vSU) [2]. In the first two phases of the StandAV in particular, the vSU is associated with uncertainties due to the process-related increasing data availability. Due to the complexity of the phenomena and processes considered and the temporal and spatial scale, uncertainties are unavoidable. Both the ordinance of safety requirements for the final disposal of high-level radioactive waste (EndlSiAnfV) [3] and the ordinance of preliminary safety investigations in the site selection procedure (EndlSiUntV) [2] require the investigation of the uncertainties but do not specify how the uncertainties should be dealt with methodically. An evaluation based on methodological approaches helps to make the process transparent and comprehensible.

The BewUSt research project with a focus on uncertainties is carried out at BASE since November 2023. The main objectives of the project are, on the one hand, to create an overview of possible uncertainties for all host rocks to be considered in Germany and, on the other hand, to determine which qualitative and quantitative methodological approaches have been used to deal with uncertainties in recent relevant projects. In addition, the BewUSt research project will prepare the results in a way that they are appropriate to communicating uncertainties for specific target groups.

As part of the project tasks, the uncertainties are therefore characterized and existing methodological approaches are identified. In addition, it is examined whether approaches from other areas can be used. Furthermore, the current status of common methodological approaches (e.g. sensitivity analysis) as well as unconventional probabilistic methods from the field of final disposal and other geoscientific fields (e.g. hydrogeology or reservoir modelling) are determined and their application to the uncertainties relevant to vSU are examined. The poster shows exemplary work results e.g. for potential uncertainties and their methodical treatment. The results will help BASE to independently examine and evaluate the proposals of the implementer.

[1] Deutscher Bundestag (BT) (2020): Law on the Search and Selection of a Site for a Final Repository for High-Level Radioactive Waste, Site Selection Act - StandAG. (07.12.2020) (BGBl. I, 26).

[2] EndlSiUntV (2020): Ordinance on requirements for the implementation of preliminary safety investigations in the site selection procedure for the final disposal of high-level radioactive waste (Endlagersicherheitsuntersuchungsverordnung - EndlSiUntV) (6. 10. 2020) (BGBl. I S. 2094, 2103) last modified 2020 (Bundesgesetzblatt (BGBl))

[3] EndlSiAnfV (2020): Ordinance on Safety Requirements for the Final Disposal of Highly Radioactive Waste (Endlagersicherheitsanforderungsverordnung - EndlSiAnfV) (6. 10. 2020) (BGBl. I S. 2094), last modified 2020 (Bundesgesetzblatt (BGBl))

How to cite: Frieling, G., Beuth, T., and Kähler, E.: Assessment and Communication of Uncertainties in the Site Selection Procedure (BewUSt), Third interdisciplinary research symposium on the safety of nuclear disposal practices, Berlin, Germany, 17–19 Sep 2025, safeND2025-75, https://doi.org/10.5194/safend2025-75, 2025.