Deep Underground Repository Research Articles

Two-phase reactive transport modeling of heterogeneous gas production in a low- and intermediate-level waste repository

Abstract. The widely proposed approach to dispose of low- and intermediate-level radioactive waste is to store it in a deep underground repository with multiple barriers. A typical gallery is filled with concrete containers accommodating cemented steel drums of waste. Inside a container, different gases may be produced by (bio)chemical reactions, which include pH-dependent anoxic corrosion of metals and the degradation of organic matter. Both reactions consume water and may lead to pressure buildup and transport of gas, both within and around the repository. In order to investigate the controlling factors of this gas production process, a coupled reactive transport model of component-based two-phase flow in the OpenGeoSys framework is adopted here. The numerical study of Huang et al. (2021) has shown that a realistic internal structure of a waste package, including the heterogeneous distribution of materials with different chemical and hydrological properties, and the exchange of mass at the boundaries are key factors that determine the evolution of the waste package. Based on the study of Huang et al. (2021), the geometric configuration of the model has been further extended to reflect the various conditions of a multi-container disposal in a gallery. In a two-dimensional setup several model scenarios have been designed and simulated to study the change in gas production rate over time in relation to water availability in various geological and waste storage setups. In this presentation, we show simulation results covering the geochemical evolution of a waste package over 500 years. It is found that the initial water content in the waste compartment only controls the gas production rate for the first 40 to 60 years. The early pressure buildup and gas release rates are largely controlled by several critical parameters, including the permeability of cement material and water availability at the boundary. The sensitivity of these parameters is currently being investigated in detail.

Experimental investigation of uranium release from spent nuclear fuels under conditions expected in a deep geological repository

Abstract. Direct disposal of spent nuclear fuel (SNF) in a deep geological repository is planned for the long-term management of the highly radioactive waste discharged from the light water reactors (LWRs) of nuclear power plants in Germany. The aim is to safely isolate the radionuclide inventory from the biosphere, ad infinitum, making use of a multi-barrier system consisting of technical, geotechnical and geological barriers interposed between the emplaced waste and the environment. Still, reversibility measures including retrievability of the waste products are considered in this concept. Internationally, in safety assessments for deep underground repositories (DGRs), the access of groundwater, followed by failure of canisters and finally loss of the SNF's cladding integrity is considered in the long term. Hence, the evaluation of the performance of SNF in the near field of a deep geological repository requires process understanding of SNF dissolution and rates, in addition to the quantification of the radionuclide release from SNF under reducing conditions of a breached container. With the aim to derive a radionuclide source term, the SNF dissolution and alteration processes can be assigned to two steps, namely the (i) instantaneous release of radionuclides upon cladding failure from gap and grain boundaries and (ii) a long-term release that results from the dissolution of the fuel grains themselves (Ewing, 2015). Since most radionuclides produced during irradiation of nuclear fuel in a reactor are trapped within the SNF matrix, understanding the processes leading to the dissolution of the matrix and its dissolution rates are of particular interest for post-closure safety assessments (Pastina and LaVerne, 2021). Research at the Institute for Nuclear Waste Disposal (INE) at the Karlsruhe Institute of Technology (KIT) has focused, for more than 25 years, on the behaviour of SNF (irradiated UO2 and mixed-oxide, MOX, fuels) under geochemical conditions (pH, redox milieu, gas atmosphere and ionic strength) representative of various repository concepts and radiolytic reactions on SNF dissolution mechanisms. This contribution will give an overview on the radionuclide release from dissolving SNFs under repository conditions. In particular, uranium released into solution from UO2 and MOX fuels during up to 1629 d of leaching in salt brines, granitic/bentonitic or cementitious groundwaters under Ar or Ar/H2 atmospheres. As observed in other published studies on SNF dissolution under reducing conditions, a large scatter of the uranium concentration is seen in the initial stages of our experiments. After about 1 year of leaching, the aqueous concentration of uranium slowly approaches the solubility limit of U(IV), independent of the type of studied SNF samples (various irradiated UO2 and MOX fuels). A similar behaviour is observed for other redox-sensitive actinides. The release rate of uranium decreases significantly in all experiments within first 400 d of leaching. Although the rate is very low, it is >0, and a continuous release is observed. A comparison of the results from SNF experiments under various redox conditions demonstrates that oxidative matrix dissolution is inhibited due to the presence of hydrogen. Our results are discussed in the context of the outcome of international research programmes on SNF corrosion. These findings are considered to be of relevance for the definition of source terms for various irradiated LWR fuels in safety investigations for a DGR in Germany.

A numerical study on THM coupled behavior in the high-level radioactive waste disposal system

It is essential to securely isolate high-level radioactive waste from the biosphere, and geologic disposal of it at a deep underground repository is considered the most effective method. Therefore, it is crucial to research the complex thermo-hydro-mechanical-chemical (THMC) coupled behavior in geological disposal systems parallel with numerical simulation. DECOVALEX is an international cooperating project to efficiently develop numerical methods and models and validate through test results for predicting the THMC interactions in the disposal systems. In Task C of DECOVALEX-2023, the modeling teams focus on understanding pore pressure development and THM interactions in the host rock and buffer material during the FE experiment, a 1:1 scale in-situ heater test based on Nagra’s Reference repository design. We used OGS-FLAC3D for the numerical simulation, combining OpenGeoSys for TH simulation and FLAC3D for M simulation. At the first phase of the task, a simple two-dimensional benchmark problem was defined to set up the numerical model. THM coupled processes in the bentonite were simulated with a two-phase flow system, and we investigated the temperature and pressure variations on the given monitoring position. Vaporization induced by a temperature increase in the bentonite was observed at the heater’s vicinity, and flow occurred by capillarity, and pressure difference was well simulated. Additionally, the flow process was dominant in the near field of engineering barrier, while we observed thermal pressurization in the far-field area. We plan to apply the developed model to a full-scale three-dimensional numerical simulation for the next phase.

Mechanistic understanding of Curium(III) sorption on natural K-feldspar surfaces

To assess a reliable safety case for future deep underground repositories for highly active nuclear waste the retention of radionuclides by the surrounding host rock must be understood comprehensively. Retention is influenced by several parameters such as mineral heterogeneity and surface roughness, as well as pore water chemistry (e.g., pH). However, the interplay between those parameters is not yet well understood. Therefore, we present a correlative spectromicroscopic approach to investigate sorption of the actinide Cm(III) on: 1) bulk K-feldspar crystals to determine the effect of surface roughness and pH (5.5 and 6.9) and 2) a large feldspar grain as part of a complex crystalline rock system to observe how sorption is influenced by the surrounding heterogeneous mineralogy. Our findings show that rougher K-feldspar surfaces exhibit increased Cm(III) uptake and stronger complexation. Similarly, increasing pH leads to higher surface loading and stronger Cm(III) binding to the surface. Within a heterogeneous mineralogical system sorption is further affected by neighboring mineral dissolution and competitive sorption between mineral phases such as mica and feldspar. The obtained results express a need for investigating relevant processes on multiple scales of dimension and complexity to better understand trivalent radionuclide retention by a potential repository host rock.

Twenty years of research on spent nuclear fuel in the context of final disposal at KIT-INE in multinational collaborative projects

Abstract. Disposal of spent nuclear fuel (SNF) in deep geological repositories is considered a preferential option for the management of such wastes in many countries with nuclear power plants. With the aim to permanently and safely isolate the radionuclide inventory from the biosphere for a sufficient time, a multibarrier system consisting of technical, geotechnical and geological barriers is interposed between the emplaced waste and the environment. In safety assessments for deep underground repositories, access of water, followed by failure of canisters and finally loss of the cladding integrity is considered in the long-term. Hence, evaluating the performance of SNF in deep geological disposal systems requires process understanding of SNF dissolution and rates as well as quantification of radionuclides release from SNF under reducing conditions of a breached container. In order to derive a radionuclide source term, the SNF dissolution and alteration processes can be assigned to two steps: (i) instantaneous release of radionuclides upon cladding failure from gap and grain boundaries and (ii) a long-term release that results from dissolution of the fuel grains itself (Ewing, 2015). In this context, research at KIT-INE has focused for more than 20 years on the behavior of SNF (irradiated UO2 and MOX fuels) under geochemical conditions (pH, redox and ionic strength) representative of various repository concepts, including the interaction of SNF with backfill material, such as bentonite as well as the influence of iron corrosion products, e.g. magnetite and radiolytic reactions on SNF dissolution mechanisms. Since 2001, KIT-INE has contributed with experimental and theoretical studies on the behavior of SNF under repository relevant conditions to six Euratom projects viz SFS (2001–2004), NF-PRO (2004–2006), MICADO (2006–2009), RECOSY (2007–2011), FIRST-Nuclides (2012–2014) and DISCO (2016–2021). Moreover, since 2007, overall 4 consecutive projects for the Belgian waste management organization, ONDRAF-NIRAS, were performed on the behavior of SNF under conditions representative of the Belgian “Supercontainer” concept. In this contribution, we summarize major achievements of theses research projects to understand and quantify the radionuclide release from dissolving SNF under repository conditions. In particular, the dependence of radionuclide release on the chemical composition of the aqueous and gaseous phase in the proximity of repositories in different types of host rock is discussed.

A Leaching Model for “Aged” Sodium–Aluminophosphate Vitreous Waste Forms of Radionuclides

High-level radioactive wastes of nuclear power engineering are vitrified for removal to deep underground repositories. Radionuclide decay leads to heating and crystallization of the vitrified wastes with deterioration of their ability to retain hazardous components. The model for vitrified Na–Al–P glass dissolution in water has been developed taking into account the experimental data. The objective of the model is to predict the behavior of the waste form in the repository.

Dry storage modeling activities at PSI: implementation and testing of a creep model for dry storage

Abstract According to the Swiss waste management policy, spent nuclear fuel is planned to be disposed in a deep underground repository. Prior to final disposal, spent nuclear fuel is stored at reactor sites and in a centralized dry storage facility. Since the operation dates of the final repository are unknown, extended periods of interim storage have to be considered. A research program to investigate the fuel rod integrity during long-term dry storage has recently been launched at Paul Scherrer Institute. In the context of the project, fuel rod performance simulations will be carried out with the code Falcon. Until now, an extensive literature survey concerning current trends in dry storage modelling has been written and first developments of Falcon’s capabilities towards dry storage modeling have been made. In this work, Falcon has been modified by implementing a long-term cladding creep model. The original and upgraded versions of Falcon have been used to simulate a demonstration case consisting of the base irradiation, wet storage, drying and dry storage. Results obtained with both versions have shown an important difference in the cladding hoop creep which justifies implementation of the new creep model.

A non-linear creep constitutive model for salt rock based on fractional derivatives

The creep property of the salt rock is an important determinant of deep underground repository of energy storage for long-term operatio. Starting from fractional derivative, this paper proposes a new concept of super Abel dashpot based on Abel dashpot. According to the Nishihara model, the non-linear element of super Abel dashpot is introduced, and a new salt rock creep constitutive model is established. The theoretical analytic solution of this model is then deduced. Fitting analysis of the experimental data indicates that the model can well reflect the whole-process creep curve, especially the non-linear accelerated creep stage.

Hydromechanical modelling of the SEALEX experiments

Numerical modelling of coupled physical processes in bentonite–sand mixtures under the geological conditions is significant for designing and constructing sealing systems in deep underground repositories for highly radioactive nuclear waste. Within the framework of DECOVALEX 2015, Task A, this work presents the model validation of OpenGeoSys by numerical modelling of coupled hydromechanical (HM) processes in bentonite–sand mixtures. Parameters used in the HM model were determined by modelling the laboratory tests of the sealing experiment (SEALEX). Afterwards these parameters were applied for the modelling of a small-scale mock-up test considering the influence of technological gap and incidental fail of the seal in the sealing system. In order to investigate the availability of employing these HM parameters and numerical models directly to field predictions, the modelling results and measured data of an in situ SEALEX experiment were analysed comparatively. The modelling results reproduced well the main features in HM behaviour of the compacted bentonite–sand mixture, which denotes that the adopted HM models and parameters are adequate for describing the HM processes in the sealing system. It is necessary to take the elastoplastic behaviour and evolution of the permeability of bentonite–sand mixtures into account when using the adopted models to reproduce the HM processes of a sealing system.

Potential for microbial H2 and metal transformations associated with novel bacteria and archaea in deep terrestrial subsurface sediments

Geological sequestration in deep underground repositories is the prevailing proposed route for radioactive waste disposal. After the disposal of radioactive waste in the subsurface, H2 may be produced by corrosion of steel and, ultimately, radionuclides will be exposed to the surrounding environment. To evaluate the potential for microbial activities to impact disposal systems, we explored the microbial community structure and metabolic functions of a sediment-hosted ecosystem at the Horonobe Underground Research Laboratory, Hokkaido, Japan. Overall, we found that the ecosystem hosted organisms from diverse lineages, including many from the phyla that lack isolated representatives. The majority of organisms can metabolize H2, often via oxidative [NiFe] hydrogenases or electron-bifurcating [FeFe] hydrogenases that enable ferredoxin-based pathways, including the ion motive Rnf complex. Many organisms implicated in H2 metabolism are also predicted to catalyze carbon, nitrogen, iron and sulfur transformations. Notably, iron-based metabolism is predicted in a novel lineage of Actinobacteria and in a putative methane-oxidizing ANME-2d archaeon. We infer an ecological model that links microorganisms to sediment-derived resources and predict potential impacts of microbial activity on H2 consumption and retardation of radionuclide migration.

Geologic environments for nuclear waste repositories

High-level radioactive waste (HLW) results from spent reactor fuel and reprocessed nuclear material. Since 1957 the scientific consensus is that deep geologic disposal constitutes the safest means for isolating HLW for long timescales. Nuclear power is becoming significant for the Arab Gulf countries as a way to diversify energy sources and drive economic developments. Hence, it is of interest to the UAE to examine the geologic environments currently considered internationally to guide site selection. Sweden and Finland are proceeding with deep underground repositories mined in bedrock at depths of 500m, and 400m, respectively. Equally, Canada’s proposals are deep burial in the plutonic rock masses of the Canadian Shield. Denmark and Switzerland are considering disposal of their relative small quantities of HLW into crystalline basement rocks through boreholes at depths of 5,000m. In USA, the potential repository at Yucca Mountain, Nevada lies at a depth of 300m in unsaturated layers of welded volcanic tuffs. Disposal of low and intermediate-level radioactive wastes, as well as the German HLW repository favour structurally-sound layered salt stata and domes. Our article provides a comprehensive review of the current concepts regarding HLW disposal together with some preliminary analysis of potentially appropriate geologic environments in the UAE.

Poroelasticity of the Callovo–Oxfordian Claystone

This work is devoted to an experimental investigation of the poroelastic behavior of the Callovo–Oxfordian claystone, a potential host rock for the deep underground repository of high-level radioactive waste in France. Drained, undrained, pore pressure loading and unjacketed tests were carried out in a specially designed isotropic compression cell to determine the poroelastic parameters of fully saturated specimens. Great care was devoted to the saturation procedure, and small loading rates were used to ensure full drainage conditions in drained and pore pressure tests (0.5 kPa/min) and in the unjacketed test (2 kPa/min). High-precision strain measurements were performed by ensuring direct contact between the LVDT stems and the specimen. An analysis in the framework of transverse isotropic poroelasticity provided the Biot effective stress coefficients b1 (perpendicular to bedding) between 0.85 and 0.87 and b2 (parallel to bedding) between 0.90 and 0.98 under different stress conditions (pore pressure 4 MPa, total isotropic stresses of 14 and 12 MPa, respectively). A set of equivalent isotropic poroelastic parameters was also determined and a very good compatibility between the results of different tests was found, giving confidence in the parameters determined. The unjacketed test provided a directly reliable measurement of the unjacketed modulus (Ks = 21.7 GPa) that was afterward confirmed by an indirect evaluation that showed the non-dependency of Ks with respect to the stress level. These parameters were obtained for specimens cored and trimmed in the laboratory. A parametric study was then conducted so as to provide an estimation of the parameters in situ, i.e., not submitted to the damage supported by laboratory specimens. A minimal value b = 0.77 seems to be a reasonable lower bound for the equivalent isotropic Biot parameter.

Modelling an in-situ ventilation test in the Andra Underground Research Facilities

Wastes resulting from the nuclear electricity production have to be isolated from the biosphere for a very long period of time. For this purpose, deep underground repository in weak permeable geological layers is considered as a reliable solution for the nuclear waste storage. It is however well established that during excavation, the underground drilling process engenders cracks and eventually fractures [1] that deteriorate the hydro-mechanical properties of the surrounding host material in the so-called Excavation Damaged Zone (EDZ). The EDZ behaviour is a major issue because it may constitute a preferential flow path for radionuclide migration. Consequently, the characterisation of the material transport properties and of the transfer kinetics that occur around galleries still need to be investigated. The EDZ properties may be also affected by host rock-gallery air interactions. Ventilation induced drying may also provoke additional cracking, which potentially alters the transport properties of the damaged zone. Large-scale air ventilation experiments are performed in Underground Research Laboratories (URL) that have been constructed to check the feasibility of the repository. A numerical modelling of the SDZ air ventilation test (Andra URL) performed in a low permeability rock is proposed in order to both predict the development of the EDZ during excavation and study the air interaction with the host formation during maintenance phases.

Study of stainless steel corrosion by hydrogen measurement under deoxygenated, low-temperature and basic repository conditions

Spent fuel assemblies comprising Zircaloy and stainless steel pieces are expected to be disposed of in a deep underground repository. Corrosion is considered to be a leaching source of activated radionuclides such as C-14. Japanese safety assessment has estimated the corrosion rate of stainless steel as 2 × 10−2 μm a−1 based on limited experimental results at 308 K in the presence of 3200 ppm chloride. This corrosion rate corresponds to a leaching period of 8500 years.In the present study, corrosion tests of 18Cr–8Ni austenitic stainless steel were performed under assumed disposal conditions (dilute NaOH solution, pH 12.5, 303 K, Ar environment) by continuous measurement of gaseous hydrogen over 6.5 years through use of the gas flow system. The cumulative amount of hydrogen shows a good parabolic relationship against time until one year, followed by an approximately linear relationship. This suggests that the corrosion kinetics changes from an initial diffusion process to a linear rate law that holds for a long time. Assuming a corrosion reaction of 3Fe + 4H2O → Fe3O4 + 4H2, the corrosion rate decreases with time but after one year the rate remains constant at approximately 4 × 10−4 μm a−1 over the 6 next years.

Study of thermo-hydro-mechanical processes at a potential site of an Indian nuclear waste repository

A detailed scientific study is required for the disposal of high-level radioactive wastes because they generate extremely high heat during their half-life period. Although, several methods have been proposed for the disposal of nuclear wastes, deep underground repository is considered to be a suitable option.In this paper, field investigation has been done near to Bhima basin of peninsular India. Detailed fracture analysis near the borehole shows very prominent maxima of fractures striking N55∘E coinciding with the trace of master basement cover metasediment fault. Physico-mechanical properties of rocks have been determined in the laboratory.The host rock chosen is granite and engineered barrier near the canister is proposed to be clay. A thermo-hydro-mechanical (THM) analysis has been done to study the effect of heat on deformations, stresses and pore-pressure variation in granite and clay barriers. For this purpose, finite difference method has been used. Suitable rheological models have been used to model elastic canister and elasto-plastic engineered barrier and host rock. It has been found that both temperature and stresses at any point in the rockmass is below the design criteria which are 100∘C for temperature and 0.2 for damage number.

Discrete element modelling of conceptual deep geological repository for high-level nuclear waste disposal

The long-lived high-level spent nuclear has to be isolated from environment for the protection of ecosystem. One of the methods suggested to isolate the nuclear waste from ecosystem is its burial in deep underground repository. In this paper, discrete element method is used to disposal of spent fuel on stability of underground space and its surrounding rock strata. Effect of temperature increment on stresses-strains and temperature variation of surrounding rockmass due to heat generated by nuclear waste is studied and discussed. Simulation was performed on both strong and jointed granite rock in which tunnel is excavated. Bentonite is used as buffer because of its high sorptivity, longevity and low permeability. It has been found that both temperature and stresses at any point in the rock mass is below the design criteria which are 100 °C for temperature over a time period of 16 years.

Selenium redox speciation and coordination in high-burnup UO2 fuel: Consequences for the release of 79Se in a deep underground repository

The chemical state of Se in high-burnup UO2 spent nuclear fuel (SNF) was investigated by micro X-ray absorption near-edge structure (μ-XANES) spectroscopy. The data were first evaluated in terms of linear combination fits of reference compounds. In a second step, the XANES data were fitted to theoretical (ab initio) spectra via geometrical optimization of atomic clusters around a central Se atom, using the FDMNES and FitIt software packages. Best fits were obtained assuming substitution of Se in occupied or vacant oxygen sites within the UO2 lattice, with 20–25% local expansion. Based on these results and chemical arguments we argue that the long-lived fission product 79Se may be stabilized to sparingly soluble Se(−II) in SNF. This would explain the failure to detect dissolved Se in SNF leaching experiments.

Licensing process characteristics of Small Modular Reactors and spent nuclear fuel repository

This paper aims to increase the understanding of the licensing processes characteristics of Small Modular Reactors (SMR) compared with licensing of spent nuclear fuel repository. The basis of the SMR licensing process development lies in licensing processes used in Finland, France, the UK, Canada and the USA. These countries have been selected for this study because of their various licensing processes and recent actions in the new NPP construction. Certain aspects of the aviation industry licensing process have also been studied and selected practices have been investigated as possibly suitable for use in nuclear licensing.Suitable features for SMR licensing are emphasized and suggested. The licensing features of the spent nuclear fuel deep repository along with similar features of SMR licensing are discussed. Since there are similar types of challenges of lengthy licensing time frames, as well as modular features to be taken into account in licensing, these two different nuclear industry fields can be compared.The main SMR features to take into account in licensing are:•Standardization of the design.•Modularity.•Mass production.•Serial construction.Modularity can be divided into two different categories: the first category is simply a single power plant unit constructed of independently engineered modules (e.g. construction process for Westinghouse AP-1000 NPP) and the second one a power plant composed of many reactor modules, which are manufactured in factories and installed as needed (e.g. NuScale Power SMR design).The deep underground repository for spent nuclear fuel, such as the one planned to be built to Olkiluoto, Finland, is based on modular design and phased construction. In this case the disposal tunnels, bentonite buffer and backfill and the disposal canisters can be considered as modules. Due to its phased construction the facility has similar licensing characteristics of SMRs.A common licensing process needs to be simple and clear to ensure the feasibility of construction and deployment of SMR's as well as the deep geological repositories.One of the important aspects of nuclear licensing is the compliance to licensing requirements. In a discrete and transparent matter, systems engineering (SE) and requirements engineering (RE) are discussed with the licensing development. The sheer number of licensing requirements is daunting, interconnected in a complex fashion and long-lasting. The process and tools play a critical role in the success of the licensing effort.A few proposed examples of combined SMR and deep repository licensing applications are discussed using SE and RE principles. A synergistic discussion on licensing models, nuclear licensing simplification, as well as international approach in licensing, will be needed in the future. This would be a natural continuation to the collaborative efforts in Europe underway for existing and new large reactors.

Recrystallization of 223Ra with barium sulfate

In this work, the kinetics of barium sulfate recrystallization has been studied in acidic 0.01 mol dm -3 sodium sulfate solution using Ra-223 and Ba-133 tracers at very low total radium concentration, i. e. less than 10(-13) mol dm(-3). It was found that the system follows the homogeneous recrystallization model and that recrystallization rates, inferred by the decrease of Ra-223 and Ba-133 in the aqueous solution, are fast. Therefore, even at very low concentrations, below the solubility limit, radium will be retained by barium sulfate-a mineral present in the deep underground repository.

Characterization of gas flow through low-permeability claystone: laboratory experiments and two-phase flow analyses

Abstract For the characterization of gas migration through a low-permeability clay host rock for deep underground repositories, a comprehensive understanding of the relevant phenomena of gas and fluid flow through low-permeability clay is required. The National Cooperative for the Disposal of Radioactive Waste (Nagra) in Switzerland has developed a comprehensive programme to characterize gas flow in low-permeability Opalinus Clay through laboratory tests and detailed numerical analyses for developing appropriate constitutive models. Laboratory tests were performed on cores by two different laboratories, the Laboratory for Soil Mechanics at EPFL and the Department of Geotechnical Engineering and Geosciences at UPC. Loading tests were performed by both laboratories to study rock compressibility at different stress levels and water permeability dependence on void ratio. The water retention behaviour demonstrated by EPFL and UPC produced comparable results. Water permeability tests and fast controlled-volume air injection experiments were performed in a triaxial cell under isotropic stress conditions on two samples with flow parallel and normal to the bedding planes. A confining stress of 15 MPa was applied during gas testing, corresponding to a lithostatic pressure at a depth of c. 600 m below ground. For detailed analyses, the two-phase flow code TOUGH2 ( Pruess et al. 1999 ) was used. This considers fluid flow in both liquid and gas phases under the influence of pressure, viscous and gravity forces, according to Darcy's law. The standard analyses could not reproduce the measured pressure responses well, and the calibrated hydraulic and two-phase parameters were not consistent with the preceding water test and laboratory analyses. Implementing the non-linear behaviour in terms of the observed relationship between changes in void ratio and associated changes in permeability under different stress conditions significantly improved the simulated results, resulting in a conceptual model that well reproduced the observed injection pressure and outflow responses for both tests, parallel and normal to bedding, using a consistent parameter set.