Repository Performance Assessment Research Articles

Coupled Processes at Micro- and Macroscopic Levels for Long-Term Performance Assessment Studies of Nuclear Waste Repositories

Performance assessment of nuclear waste repositories requires state-of-the-art knowledge of radionuclide transport properties. Additionally, the short-term development under thermal pulses and the long-term development of the near field—due to influences such as gas generation—must be evaluated. Key thermal-hydro-mechanical-chemical processes are strongly coupled on different spatial and temporal scales. To understand these coupling mechanisms, numerous material models and numerical codes have been developed. However, the existing constitutive approaches—which have been adapted to describe small-scale laboratory experiments and validated against real-scale field observations—are often unable to capture long-term material behavior with sufficient precision. To build the confidence, a more comprehensive understanding of the system at micro- and macroscopic scales is required. Most observed macroscopic processes result from microscopic changes in the crystal structure and/or crystalline aggregates, as well as changes in material properties under the influence of various factors. To characterize these physical fields in crystals, microscopic investigations, such as visualization, or geophysical methods are introduced to verify the understanding at the microscale. Two cases are demonstrated for the presented concept using microscale information: one deals with the mechanically and thermally driven migration of fluid inclusions in rock salt, the other with dilatancy-controlled gas transport in water-saturated clay material.

Review of secondary phases formed under natural alkaline conditions at low temperatures and implications for cement–bentonite interactions in radioactive waste repositories

Abstract The alteration of bentonite under alkaline conditions and the subsequent changes in properties such as permeability and self-sealing ability should be evaluated for the performance assessment of radioactive waste repositories. As the period of evaluation for alterations is extremely long, natural analog (NA) studies that can observe long-term phenomena similar to the system of radioactive waste repositories have significant advantages. However, locations that can be set up as NA study sites with significant similarity are limited and should be pursued by localizing analog systems. This literature review summarized studies reporting secondary phases formed at low temperatures (<100°C) under a broader range of natural sites which are chemically similar to the alkaline conditions expected at bentonite in radioactive waste repositories, including near pyrometamorphic rocks, near ophiolites, and in alkaline saline lakes. This review provides insights into the species, formation conditions, and stability of secondary phases that could be formed during cement–bentonite interactions and the timescale for mineralogical transitions from a metastable to a stable phase. The findings could be useful for selecting secondary phases to be considered in reactive transport modeling for predicting cement–bentonite interactions in radioactive waste repositories.

Performance assessment of a generic repository in salt host rock based on Task F of DECOVALEX-2023 – BASE's lessons learnt and future work

Abstract. The evaluation of coupled effects of mechanical deformation, fluid and gas flow, thermal loading, and chemical reactions is an important element of the performance and safety assessment of disposal systems for high-level radioactive waste and spent nuclear fuel in deep geological formations. Early investigations showed that to be able to conduct such an evaluation there was a need to enhance process understanding and to develop models capable of simulating coupled thermal–hydro–chemical–mechanical processes over relevant timescales and spatial scales. Recognising these challenges and needs, the DECOVALEX (DEvelopment of COupled models and their VALidation against Experiments) (https://decovalex.org, last access: 24 July 2023) was initiated in 1992 as an international cooperative project of nuclear waste organisations. BASE is one of the funding organisations for the current phase of DECOVALEX-2023 (running from 2020 to 2023) and actively contributes to Task F of the project (https://www.base.bund.de/DE/themen/fa/soa/projekte-aktuell/projekte-aktuell_node.html, last access: 24 July 2023, – “Beteiligung an der internationalen Forschungskooperation DECOVALEX”). Task F (https://www.decovalex.org/D-2023/task-f.html, last access: 24 July 2023) of DECOVALEX seeks to build confidence in the models, methods, and software used for performance assessment of deep geologic repositories and to highlight additional research and development needed to improve performance assessment methodologies. The task focuses on two illustrative disposal concepts for spent nuclear fuel: a generic repository in a fractured crystalline host rock and a generic repository in a salt formation. For each of the disposal concepts a reference case is defined that includes a set of conceptual models and parameters describing features, events, and processes that impact repository performance. Modelling teams work independently and are responsible for determining how best to implement and couple the models. Each team develops its own model(s), calculates agreed performance indicator values, and performs uncertainty and sensitivity analysis. The comparison of model results is conducted in stages, commencing with a comparison of key outputs of individual process models, followed by a comparison of a single deterministic simulation of the full reference case, and proceeding to uncertainty and sensitivity analysis. The present contribution describes recent BASE activities within Task F Salt of DECOVALEX-2023. Lessons learnt are discussed and possible future steps are presented.

Newton trust-region methods with primary variable switching for simulating high temperature multiphase porous media flow

Coupling multiphase flow with energy transport due to high temperature heat sources introduces significant new challenges since boiling and condensation processes can lead to dry-out conditions with subsequent re-wetting. The transition between two-phase and single-phase behavior can require changes to the primary dependent variables adding discontinuities as well as extending constitutive nonlinear relations to extreme physical conditions. Practical simulations of large-scale engineered domains lead to Jacobian systems with a very large number of unknowns that must be solved efficiently using iterative methods in parallel on high-performance computers. Performance assessment of potential nuclear repositories, carbon sequestration sites and geothermal reservoirs can require numerous Monte-Carlo simulations to explore uncertainty in material properties, boundary conditions, and failure scenarios. Due to the numerical challenges, standard NR iteration may not converge over the range of required simulations and require more sophisticated optimization method like trust-region.We use the open-source simulator PFLOTRAN for the important practical problem of the safety assessment of future nuclear waste repositories in the U.S. DOE geologic disposal safety assessment Framework. The simulator applies the PETSc parallel framework and a backward Euler, finite volume discretization. We demonstrate failure of the conventional NR method and the success of trust-region modifications to Newton’s method for a series of test problems of increasing complexity. Trust-region methods essentially modify the Newton step size and direction under some circumstances where the standard NR iteration can cause the solution to diverge or oscillate. We show how the Newton Trust-Region method can be adapted for Primary Variable Switching (PVS) when the multiphase state changes due to boiling or condensation. The simulations with high-temperature heat sources which led to extreme nonlinear processes with many state changes in the domain did not converge with NR, but they do complete successfully with the trust-region methods modified for PVS. This implementation effectively decreased weeks of simulation time needing manual adjustments to complete a simulation down to a day. Furthermore, we show the strong scalability of the methods on a single node and multiple nodes in an HPC cluster.

Reduced-order modeling of near-field THMC coupled processes for nuclear waste repositories in shale

Performance assessment (PA) of geologic radioactive waste repositories requires three-dimensional simulation of highly nonlinear, thermo-hydro-mechanical-chemical (THMC), multiphase flow and transport processes across many kilometers and over tens to hundreds of thousands of years. Integrating the effects of a near-field geomechanical process (i.e. buffer swelling) into coupled THC simulations through reduced-order modeling, rather than through fully coupled geomechanics, can reduce the dimensionality of the problem and improve computational efficiency. In this study, PFLOTRAN simulations model a single waste package in a shale host rock repository, where re-saturation of a bentonite buffer causes the buffer to swell and exert stress on a highly fractured disturbed rock zone (DRZ). Three types of stress-dependent permeability functions (exponential, modified cubic, and Two-part Hooke’s law models) are implemented to describe mechanical characteristics of the system. Our modeling study suggests that compressing fractures reduces DRZ permeability, which could influence the rate of radionuclide transport and exchange with corrosive species in host rock groundwater that could accelerate waste package degradation. Less permeable shale host rock delays buffer swelling, consequently retarding DRZ permeability reduction as well as chemical transport within the barrier system.

Iodine effective diffusion coefficients through volcanic rock: Influence of iodine speciation and rock geochemistry

Accurate prediction of the subsurface transport of iodine species is important for the assessment of long-term nuclear waste repository performance, as well as monitoring compliance with the Comprehensive Nuclear-Test-Ban Treaty, given that radioiodine decays into radioxenon. However, the transport of iodine through intact geologic media is not well understood, compromising our ability to assess risk associated with radioiodine migration. The current study's goal is to quantify the matrix diffusion of iodine species through saturated volcanic rock, with particular attention paid to the redox environment and potential speciation changes. Diffusion experiments were run for iodide through lithophysae-rich lava, lithophysae-poor lava, and welded tuff, whereas iodate diffusion was studied through welded tuff. Iodine transport was compared with a conservative tracer, HDO, and effective diffusion coefficients were calculated. Likely due to a combination of size and anion exclusion effects, iodine species diffused more slowly than the conservative tracer through all rock types tested. Furthermore, oxidation of iodide to iodate was observed in the lithophysae-poor lava, affecting transport. Results provide much needed data for subsurface transport models that predict radioiodine migration from underground sources, and indicate the pressing need for geochemical and redox interactions to be incorporated into these models.

Uncertainty and sensitivity analysis of the chemistry of cesium sorption in deep geological repositories

Sorption is the most important process responsible for the retention of radionuclides or heavy metals in repositories. A major task in understanding sorption is to account for uncertainties that affect the long-term repository performance, and eventually, study their effect on the migration of radionuclides. This paper presents a novel comprehensive uncertainty and sensitivity analysis study of a detailed sorption chemistry model used in Cesium reactive transport modeling, based on the Morris and Sobol method. The analysis addresses the effects on the transport of Cesium through clay, exemplified for Cesium migration in the vicinity of a deep geological repository for nuclear waste, due to uncertainties in 1) site-specific detailed equilibrium sorption reactions constants and 2) clay pore water cations’ concentration, i.e. the pore water composition. The results show that the equilibrium sorption reaction constants on Type2 and FES sites in addition to the concentrations of the respective cations involved in these surface reactions are the most important uncertain parameters affecting the transport of Cesium. Furthermore, a classification tree is constructed to identify the combination of parameters values leading to the maximum Cesium concentrations at a specified location. Finally, a new Cesium isotherm and Kduncertainty range are constructed numerically based on our deterministic complex sorption chemistry calculations using the MCOTAC reactive transport code. These results can be used in large-scale repository performance assessment to inform the estimates for Cesium transport parameters, with values solidly based on detailed reactive transport modeling. The presented approach can be used for parameter uncertainty analysis of any complex reactive transport problem and can be applied to analyze any radionuclide of interest.

Contribution of the results of the CEBAMA project to decrease uncertainties in the Safety Case and Performance Assessment of radioactive waste repositories

Cement-based materials are key materials used in repositories of low, intermediate and high-level radioactive waste in any host-rock concept. These materials are used for waste conditioning, liners, seals as well as structural components. The CEBAMA (CEment-BAsed MAterials, properties, evolution, barrier functions) project (2015–2019) has been an initiative granted by the European Commission under the Horizon 2020 Research and Training Programme of the European Atomic Energy Community (EURATOM), focused on the study of cementitious materials. The project has addressed key issues of relevance for long-term safety and key scientific questions related to the use of cement-based materials in nuclear waste disposal applications. This publication presents the advances resulting from the research undertaken in the CEBAMA project, with special emphasis on their contribution to decrease uncertainties in the Safety Case and the Performance Assessment of radioactive waste repositories. The analysis is presented in three different axes, in coincidence with the lines of research of the project: degradation of cement-based materials and its impact on cement/clay interfaces; retention of safety relevant radionuclides and toxicants on cement-based materials and modelling advances in cementitious related systems. The research investigations have provided important new insights on process understanding and model developments which have significantly decreased the level of uncertainty.

Reactive transport modeling of glass alteration in a fractured vitrified nuclear glass canister: From upscaling to experimental validation

Performance assessment of geological nuclear waste repositories entails modeling of the long-term evolution of the aqueous alterations of the fractured nuclear glass block, because the time scales under consideration are of several thousands of years and hence beyond the range of any direct experimental perspectives. In this study, our objective is to bridge the gap between the reservoir-scale flow and transport simulations and the micron-scale modeling of the glass-water interfacial processes by providing quantitative evaluation of the aqueous alteration of glass at the block scale. In particular, calculations of the equivalent diffusive, hydraulic, and alteration kinetics properties and reactive transport simulations are discussed. Prior to performing reactive transport modeling at the scale of the glass canister, the preferred upscaling techniques were first applied to a synthetic fracture network system with ends to compare the results of the borosilicate glass alteration with the discrete fracture modeling and the equivalent porous medium approach. The evolution of the altered glass obtained from reactive transport modeling applied to several realizations of the equivalent fracture network tessellation was compared to the experimental data of the aqueous alteration test of a nonradioactive full-scale SON68 glass canister. The proposed model agrees with the experimental data and offers, for the first time, an opportunity to better understand the impact of fracturing as the convection due to the radioactivity acting as a heating source on the corrosion of nuclear glass.

A new scripting library for modeling flow and transport in fractured rock with channel networks

Deep crystalline bedrock formations are targeted to host spent nuclear fuel owing to their overall low permeability. They are however highly heterogeneous and only a few preferential paths pertaining to a small set of dominant rock fractures usually carry most of the flow or mass fluxes, a behavior known as channeling that needs to be accounted for in the performance assessment of repositories. Channel network models have been developed and used to investigate the effect of channeling. They are usually simpler than discrete fracture networks based on rock fracture mappings and rely on idealized full or sparsely populated lattices of channels. This study reexamines the fundamental parameter structure required to describe a channel network in terms of groundwater flow and solute transport, leading to an extended description suitable for unstructured arbitrary networks of channels. An implementation of this formalism in a Python scripting library is presented and released along with this article. A new algebraic multigrid preconditioner delivers a significant speedup in the flow solution step compared to previous channel network codes. 3D visualization is readily available for verification and interpretation of the results by exporting the results to an open and free dedicated software. The new code is applied to three example cases to verify its results on full uncorrelated lattices of channels, sparsely populated percolation lattices and to exemplify the use of unstructured networks to accommodate knowledge on local rock fractures.

Coupled THMC models for bentonite in an argillite repository for nuclear waste: Illitization and its effect on swelling stress under high temperature

Subsurface manipulations such as those expected from the disposal of heat-emanating radioactive waste in deep repositories can induce strongly coupled Thermal (T), hydrological (H), mechanical (M) and chemical (C) processes. Adequate coupled THMC models are highly desirable or even indispensable for performance assessment of such repositories, for examples for the analysis of bentonite-based engineered barrier system (EBS) surrounding the emplaced waste. In this study, we present coupled THMC model simulations of a generic nuclear waste repository in an argillite with a bentonite-based buffer. The objective is to evaluate the chemical changes in the EBS bentonite and their effect on mechanical behavior under high temperature, attempting to shed light on whether EBS bentonite can sustain temperatures higher than 100°C without significant impact on barrier's performance.Two scenarios were simulated for comparison: a case in which the temperature near the waste canister peaks at 200°C and a case in which the temperature at the same spot culminate with about 100°C. Simulations for a generic case with Kunigel-VI bentonite as backfill and Opalinus Clay as host rock were conducted for 1000years and reported in the previous study (Zheng et al., 2015). In this paper, simulations for 100,000years have been done for two types of bentonite-based buffer materials: Kunigel-VI and FEBEX bentonite. This enables us to evaluate how different types of bentonite behave in terms of the illitization and its impact on swelling stress and whether we can generalize these results to support decision making. The simulations show the occurrence of illitization in the bentonite buffer and the enhancement of illitization under high temperature. However, FEBEX bentonite undergoes less illitization mainly due to the higher ion concentration in pore water and the lower content of K-feldspar in the bentonite mineral composition. Moreover, the reduction of swelling stress by chemical changes is more pronounced for Kunigel-VI bentonite than for FEBEX bentonite. Overall, the results of our model simulations suggest that an argillite repository with a bentonite-based EBS that is similar to FEBEX bentonite could sustain temperatures much higher than 100°C as far as illitization concerns. Model results also reveal that illitization is stabilized after about 2000years in bentonite near the waste package, but continues in bentonite near the bentonite-argillite interface, which manifests the strong effect of geochemical interaction between EBS bentonite and host rock on long term illitization in bentonite.

Reductive precipitation of neptunium on iron surfaces under anaerobic conditions

Reductive precipitation of the radiotoxic nuclide 237Np from nuclear waste on the surface of iron canister material at simulated deep repository conditions was investigated. Pristine polished as well as pre-corroded iron specimens were interacted in a deoxygenated solution containing 10–100 μM Np(V), with 10 mM NaCl and 2 mM NaHCO3 as background electrolytes. The reactivity of each of the two different systems was investigated by analyzing the temporal evolution of the Np concentration in the reservoir. It was observed that pre-oxidized iron specimen with a 40 μm Fe3O4 corrosion layer are considerably more reactive regarding the reduction and immobilization of aqueous Np(V) as compared to pristine polished Fe(0) surfaces. 237Np immobilized by the reactive iron surfaces was characterized by scanning electron microscopy as well as synchrotron-based micro-X-ray fluorescence and X-ray absorption spectroscopy. At the end of experiments, a 5–8 μm thick Np-rich layer was observed to be formed ontop of the Fe3O4 corrosion layer on the iron specimen. The findings from this work are significant in the context of performance assessments of deep geologic repositories using iron as high level radioactive waste (HLW) canister material and are of relevance regarding removing pollutants from contaminated soil or groundwater aquifer systems.

Fundamentals of the NEA Thermochemical Database and its influence over national nuclear programs on the performance assessment of deep geological repositories

For the last 30 years, the NEA Thermochemical Database (TDB) Project (www.oecd-nea.org/dbtdb/) has been developing a chemical thermodynamic database for elements relevant to the safety of radioactive waste repositories, providing data that are vital to support the geochemical modeling of such systems. The recommended data are selected on the basis of strict review procedures and are characterized by their consistency. The results of these efforts are freely available, and have become an international point of reference in the field. As a result, a number of important national initiatives with regard to waste management programs have used the NEA TDB as their basis, both in terms of recommended data and guidelines. In this article we describe the fundamentals and achievements of the project together with the characteristics of some databases developed in national nuclear waste disposal programs that have been influenced by the NEA TDB. We also give some insights on how this work could be seen as an approach to be used in broader areas of environmental interest.

The use of positrons to survey alteration layers on synthetic nuclear waste glasses

In order to safeguard society and the environment, understanding radioactive waste glass alteration mechanisms in interactions with solutions and near-field materials, such as Fe, is essential to nuclear waste repository performance assessments. Alteration products are formed at the surface of glasses after reaction with solution. In this study, glass altered in the presence of Fe0 in aqueous solution formed two alteration layers: one embedded with Fe closer to the surface and one without Fe found deeper in the sample. Both layers were found to be thinner than the alteration layer found in glass altered in aqueous solution only. For the first time, Doppler Broadening Positron Annihilation Spectroscopy (DB-PAS) is used to non-destructively characterize the pore structures of glass altered in the presence of Fe0. Advantages and disadvantages of DB-PAS compared to other techniques used to analyze pore structures for altered glass samples are discussed. Ultimately, DB-PAS has shown to be an excellent choice for pore structure characterization for glasses with multiple alteration layers. Monte Carlo modeling predicted positron trajectories through the layers, and helped explain DB-PAS data, which showed that the deeper alteration layer without Fe had a similar composition and pore structure to layers on glass altered in water only.

Development of performance assessment models for glass dissolution

ABSTRACTNUMO and JAEA have been conducting a joint research since FY2011, which is aimed to enhance the methodology of repository design and performance assessment in preliminary investigation stage for the deep geological disposal of high-level radioactive waste. As a part of this joint research, we have been developing glass dissolution models which include various processes derived from glass-overpack-bentonite buffer interaction, considering the precipitation of Fe-silicates associated with steel overpack corrosion, and Si transport through altered layer of glass. The objective of this modeling work is to show comprehensively the lifetime of the vitrified waste due to glass matrix dissolution timescales through sensitivity analysis, and to identify the feature/process that most strongly influences the lifetime, and to identify future R&D issues that would help to improve the nuclide transport analysis with confidential value and the safety case in future. The sensitivity analysis suggested that the duration of the glass dissolution might be predicted in the ranges from 3.8×103 to 1.9×105 years. Also, the results indicated that the precipitation of Fe–silicate has the strongest influence on the long-team behavior of vitrified waste.

Sequestration of radioactive iodine in silver-palladium phases in commercial spent nuclear fuel

Radioactive iodine is the Achilles' heel in the design for the safe geological disposal of spent uranium oxide (UO2) nuclear fuel. Furthermore, iodine's high volatility and aqueous solubility were mainly responsible for the high early doses released during the accident at Fukushima Daiichi in 2011. Studies Kienzler et al., however, have indicated that the instant release fraction (IRF) of radioiodine (131/129I) does not correlate directly with increasing fuel burn-up. In fact, there is a peak in the release of iodine at around 50–60 MW d/kgU, and with increasing burn-up, the IRF of 131/129I decreases. The reasons for this decrease have not fully been understood. We have performed microscopic analysis of chemically processed high burn-up UO2 fuel (80 MW d/kgU) and have found recalcitrant nano-particles containing, Pd, Ag, I, and Br, possibly consistent with a high pressure phase of silver iodide in the undissolved residue. It is likely that increased levels of Ag and Pd from 239Pu fission in high burnup fuels leads to the formation of these metal halides. The occurrence of these phases in UO2 nuclear fuels may reduce the impact of long-lived 129I on the repository performance assessment calculations.

Numerical modeling of colloid-facilitated radionuclide decay chain transport in a coupled fracture–matrix system

A numerical model is developed to simulate the colloid-facilitated transport of radionuclides in a single fracture-host rock coupled system with planar geometry. Such a system may be encountered in case of leaching and transport of radionuclides from a high-level radioactive waste repository in deep geological formations. The presence of colloids in the system reduces the retardation effects on the radionuclide migration and can change the concentration distribution of radionuclides in the system significantly. The processes such as advection, dispersion, surface sorption, radioactive decay and diffusive loss to the host rock are considered for transport in the fracture, whereas dispersion, adsorption and radioactive decay are included for transport in the host rock. Since the radionuclides to be disposed into the high-level waste repository may include a decay chain of progeny, the performance assessment methodology must incorporate the ingrowth of progeny with time to carry out realistic estimates. The inclusion of decay chain buildup and transport is essential to avoid underestimation with respect to the environmental impact assessment for the repository. Hence, the numerical model based on implicit finite difference scheme is developed with the inclusion of decay chain transport. The model is verified by comparing with well-known analytical solutions. The application of the model is demonstrated by applying it to a five-member decay chain of 238U. With increase in the concentration of colloids, the concentrations of the radionuclides in the fracture water increase and also reach the steady state at an earlier time. The concentrations of radionuclides almost become steady throughout the distance of 200 m in the fracture as the colloid concentration reaches 5 kg m−3 or more. Sensitivity analysis is carried out with respect to various model parameters such as dispersion coefficient in fracture, dispersion coefficient in the porous host rock, porosity of the host rock and distribution coefficient of radionuclide with respect to the mobile colloids. The developed model can be a useful tool for the performance assessment of high-level radioactive waste repositories.

Porewater chemistry in compacted bentonite: Application to the engineered buffer barrier at the Olkiluoto site

Compacted bentonite is used as sealing and buffer material in engineered barrier systems (EBS) of high-level radioactive waste repositories. The chemical characteristics of this clay and its porewater affect the migration of radionuclides eventually released from the waste. They also determine the integrity and long-term performance of the clay barriers. Key features are the structural negative charge and the large proportion of structural (interlayer) water of the main mineral montmorillonite, which leads to exclusion of anions and a surplus of cations in a large part of the porosity space. The objective of this contribution was to assess the impact of different porosity model concepts on porewater chemistry in compacted bentonite in the context of the planned Finnish spent nuclear fuel repository at Olkiluoto. First, a structural model based on well-established crystallographic and electrostatic considerations was set up to estimate the fractions of the different porosity types. In view of the uncertainty related to the chemical properties of the interlayer water, two very different model concepts (anion-free interlayer, Donnan space), together with a well-established thermodynamic model for bentonite, were applied to derive the porewater composition of the bentonite buffer at Olkiluoto. The simulations indicate very similar results in the “free” water composition for the two models and thus support the validity of the reference porewater concept commonly used in performance assessment of waste repositories. Differences between the models are evident in the composition of the water affected by the surface charge (i.e. diffuse double layer and interlayer). These reflect the conceptual uncertainty in current multi-porosity diffusion models.

Adsorption of actinides within speleothems

AbstractStalagmites and stalactites, as observed within natural caves, may develop inside geological repositories during constructional and post-operational periods. It is therefore important to understand actinide sorption within such materials. Towards this, experimental studies were carried out with233U,238Np (VI),238Np (IV),239Pu and241Am radiotracers using natural speleothem samples collected from the Dharamjali cave of the Kumaon Lesser Himalayas, India. Petrological/mineralogical studies showed that natural speleothems have three general domains: (1) columnar calcite; (2) microcrystalline calcite; and (3) botryoidal aragonite – each with ferruginous materials. Results showed that all domains of speleothems can take up >99% actinides, irrespective of valence state and pH (1–6 range) of the solution. However, distribution coefficients were found to be at a maximum in aragonite for most of the actinides. Such data are very important for long-term performance and safety assessments of the deep geological repositories planned for the disposal of high-level nuclear wastes.

The effect of stagnant water zones on retarding radionuclide transport in fractured rocks: An extension to the Channel Network Model

An essential task of performance assessment of radioactive waste repositories is to predict radionuclide release into the environment. For such a quantitative assessment, the Channel Network Model and the corresponding computer program, CHAN3D, have been used to simulate radionuclide transport in crystalline bedrocks. Recent studies suggest, however, that the model may tend to underestimate the rock retarding capability, because it ignores the presence of stagnant water zones, STWZs, situated in the fracture plane. Once considered, the STWZ can provide additional surface area over which radionuclides diffuse into the rock matrix and thereby contribute to their retardation.The main objective of this paper is to extend the Channel Network Model and its computer implementation to account for diffusion into STWZs and their adjacent rock matrices.In the first part of the paper, the overall impact of STWZs in retarding radionuclide transport is investigated through a deterministic calculation of far-field releases at Forsmark, Sweden. Over the time-scale of the repository safety assessments, radionuclide breakthrough curves are calculated for increasing STWZ width. It is shown that the presence of STWZs enhances the retardation of most long-lived radionuclides except for 36Cl and 129I.The rest of the paper is devoted to the probabilistic calculation of radionuclide transport in fractured rocks. The model that is developed for transport through a single channel is embedded into the Channel Network Model and new computer codes are provided for the CHAN3D. The program is used to (I) simulate the tracer test experiment performed at Äspö HRL, STT-1 and (II) investigate the short- and long-term effect of diffusion into STWZs. The required data for the model are obtained from detailed hydraulic tests in boreholes intersecting the rock mass where the tracer tests were made.The simulation results fairly well predict the release of the sorbing tracer 137Cs. It is found that over the short time-scale of the tracer experiment, the effect of diffusion into STWZs is not as pronounced as that of matrix diffusion directly from the flow channel, and the latter remains the main retarding mechanism. Predictions for longer time-scale, tens of years and more, show that the effect of STWZs becomes strong and tends to increase with transport time. It is shown that over the long times of interest for safety assessment of radioactive waste repositories, STWZs can substantially contribute to radionuclide retardation, though for the short time-scales the impact is not very strong and is not expected to affect the results of short-term field experiments.