Janez Perko

Decalcification of cracked cement structures

The benchmark problem presented in this paper deals with the leaching of calcium from hardened cement paste. The leaching of calcium results in the dissolution of the cement minerals which affects physical, chemical and mechanical properties of porous cement matrix. The dissolution of cement minerals in this case progresses heterogeneously as a consequence of a small-scale geometrical feature (crack) within a domain. Complexity of transport through cracked porous media combined with complex cement chemistry can lead to considerable modelling uncertainties. One possible way to get an insight into the robustness of modelling results is to perform benchmark based on (i) different transport models and solution methods (finite volume, finite element, etc.), (ii) different geochemical solvers and (iii) different coupling algorithms (sequential iterative and non-iterative). This benchmark is designed to gradually increase the complexity of the problem and in this way recognize modelling elements that are the most sensitive in terms of modelling results, e.g. evolution of physical and chemical properties. Five international teams participated in this benchmark exercise. The reactive transport codes used (HYTEC, MIN3P, OGS-GEM, ORCHESTRA, COMSOL Multiphysics-iPHREEQC) give similar patterns in terms of predicted concentrations of elements and the mineralogy. The level of agreement depends on the problem complexity related mainly to the weighting and conservation properties of different numerical methods, to the coupling between transport and reactive solver and the agreement of thermodynamic database.

Concrete in Engineered Barriers for Radioactive Waste Disposal Facilities: Phenomenological Study and Assessment of Long Term Performance

The paper aims to highlight recent developments at the Belgian Nuclear Research Centre SCK.CEN in experimental and numerical study of the coupled physical-chemical behaviour of concrete subject to chemical degradation. The discussion mainly focusses on three interlinked research projects covering novel experimental methods to study the alteration of hydraulic and transport properties during carbonation and calcium leaching, a pore scale numerical model to capture microstructural changes due to the above degradation processes and a generic multiscale model aimed at determining evolution of the properties of a macrostructure over the long term. The paper also describes supplementary continuum scale numerical studies concerning concrete-clay interactions and geochemical impact on the physical structure of concrete. Preliminary findings from these studies show encouraging results such as the development of novel leaching, water permeability and diffusion apparatus, a robust pore scale model based on Lattice-Boltzmann method and a mesoscale study focused on the importance of interfacial transition zones on the effective diffusivity for linear and nonlinear diffusion problems.

Diffusion velocity lattice Boltzmann formulation applied to transport in macroscopic porous media

This paper describes the application of a single relaxation time (SRT) lattice Boltzmann scheme to the transport in porous media with large spatial variations of diffusion coefficients. Effective diffusion coefficients can vary substantially within porous media because of their dependence on porosity and tortuosity which can span over several orders of magnitude, depending on pore size and connectivity. Moreover, when mass is transported with pore-water in porous media, the hydrodynamic dispersion, which depends on Darcys velocity, contributes additionally to the usually anisotropic variation of the dissipative term. In contrast to the traditional treatment of spatially variable diffusion coefficient by the variation of a SRT, here the variability is accommodated through the use of diffusion velocity formulation which allows for larger variabilities of diffusion coefficient. The volume averaged properties of mass transport in macroscopic porous media are resolved through the additional source term which is similar to the existing force adjusting methods. The applicability of both the proposed schemes is demonstrated on two examples. The first demonstrates that the method is accurate for the large variation of diffusion coefficients and porosities. The second example introduces mass diffusion in a real, geometrically complex system with spatially contrasting properties.

COUPLING TIME-DEPENDENT SORPTION VALUES OF DEGRADING CONCRETE WITH A RADIONUCLIDE MIGRATION MODEL

Safety assessment of radioactive waste disposal facilities is usually carried out by means of simplified models. Abstraction of the numerical model from the real physical environment is done in several steps. One of the most challenging issues in safety assessment concerns the long time scales involved and the evolution of engineered barriers over thousands of years. For some processes occurring in specific engineered barriers the uncertainties related to long time scales are addressed by implementing conservative assumptions in the radionuclide migration models. Other processes such as chemical concrete degradation, however, can be estimated for long time periods by the use of coupled geochemical transport models. For many near-surface disposal facilities, concrete is a very important engineered barrier because it is used in the construction of disposal modules or vaults, in production of high-integrity monoliths and their backfilling and for waste conditioning. Knowledge on the durability of such concrete components and its relation to radionuclide sorption is important for a defensible safety assessment. Chemical degradation typically occurs as the result of decalcification, dissolution and leaching of cement components and carbonation. These reactions induce a gradual change in the solid phase composition and the concrete pore-water composition, from “fresh” concrete porewater with a pH above 13 to a pH lower than 10 for “evolved” porewater associated with fully degraded concrete. The focus of this work is to analyse the behaviour of the disposal facility in terms of radionuclides sorption values depending on the geochemical evolution of engineered barriers. The time-dependency of the concrete mineralogy and porewater is coupled with sorption values that are characteristic for the four concrete degradation states: (i) State I with a pH larger than 12.5, controlled by the dissolution of alkali-oxides, (ii) State II with a pH at 12.5 controlled by the dissolution of portlandite, (iii) State III with a pH between 12.5 and 10 when all portlandite is dissolved and the pore water composition is determined by different cement phases including calcium-silicate hydrates (C-S-H phases), and (iv) State IV with a pH lower than 10 with calcite and aggregate minerals present. Above mentioned pH values are valid for a system with a temperature of 25°C. Sorption values are obtained from a literature review. The time-dependency of the sorption values Rd is implemented in a one-dimensional radionuclide migration model used for release calculations from the planned near-surface disposal facility at Dessel, Belgium. Calculated releases will be discussed for radionuclides typical of low- and intermediate level short-lived (LILW-SL) waste.Copyright

Evolution of Sorption Properties in Large-Scale Concrete Structures Accounting for Long-Term Physical-Chemical Concrete Degradation

Long-term safety of radioactive waste disposal facilities relies on the longevity of natural or engineered barriers designed to minimize the migration of contaminants from the facility into the environment. Especially near surface disposal facilities, such as planned by ONDRAF/NIRAS for the Dessel site in Belgium, long-term safety relies almost exclusively on the containment ability of the engineered barriers (EB) with concrete being the most important EB material used. Concrete is preferred over other materials mainly due to its favourable chemical properties resulting in a high chemical retention capacity, and owing to its good hydraulic isolation properties. However, due to the long time frames typically involved in safety assessment, the chemical, physical and mechanical properties of concrete evolve in time. The alterations in concrete mineralogy also cause changes in pH and sorption behaviour for many radionuclides during chemical degradation processes. Application of dynamic sorption of concrete requires an adequate knowledge of long-term concrete degradation processes, knowledge of the effect of changing mineralogy to sorption of radionuclides and knowledge of large-scale system behaviour over time. Moreover, when applied to safety assessment models, special attention is required to assure robustness and transparency of the implementation. The discussion in this paper focuses on the sorption properties of concrete; selection of data, rescaling issues and on the hypotheses used to build a robust and yet transparent dynamic model for large-scale concrete structures for assessing the long-term performance.© 2011 ASME

A CONSISTENT APPROACH FOR THE DEVELOPMENT OF A COMPREHENSIVE DATA BASE OF TIME-DEPENDENT PARAMETERS FOR CONCRETE ENGINEERED BARRIERS

This paper presents a consistent approach for the development of a comprehensive data base of time-dependent hydraulic and transport parameters for concrete engineered barriers of the future Dessel near surface repository for low level waste. The parameter derivation is based on integration of selected data obtained through an extensive literature review, data from experimental studies on cementitious materials specific for the Dessel repository and numerical modelling using physically-based models of water and mass transport. Best estimate parameter values for assessment calculations are derived, together with source and expert range and their probability density function wherever the data was sufficient. We further discuss a numerical method for upscaling laboratory derived parameter values to the repository scale; the resulting large-scale effective parameters are commensurate with numerical grids used in models for radionuclide migration. To accommodate different levels of conservatism in the various assessment calculations defined by ONDRAF/NIRAS, several sets of parameter values have been derived based on assumptions that introduce different degrees of conservatism. For pertinent parameters, the time evolution of such properties due to the long-term concrete degradation is also addressed. The implementation of the consistent approach is demonstrated by considering the pore water diffusion coefficient as an example.Copyright

The Long-Term Safety and Performance Analyses of the Surface Disposal Facility for the Belgian Category A Waste at Dessel

ONDRAF/NIRAS, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials, and its partners have developed long-term safety and performance analyses in the framework of the license application for a surface disposal facility for low level radioactive waste (category A waste) at Dessel, Belgium.This paper focusses on the methodology of the safety assessments and on key results from the application of this methodology.An overview is given (1) of the performance analyses for the containment safety function of the disposal system and (2) of the radiological impact analyses confirming that radiological impacts are below applicable reference values and constraints and leading to radiological criteria for the waste and the facility. In this discussion, multiple indicators for performance and safety are used to illustrate the multi-faceted nature of long-term performance and safety of the surface disposal.This contributes to the multiple lines of reasoning for confidence building that a positive decision to proceed to the next stage of construction is justified.Copyright

Influence of Cracks in Cementitious Engineered Barriers in a Near-Surface Disposal System: Assessment Analysis of the Belgian Case

In large cement-based structures such as a near surface disposal facility for radioactive waste voids and cracks are inevitable. However, the pattern and nature of cracks are very difficult to predict reliably. Cracks facilitate preferential water flow through the facility because their saturated hydraulic conductivity is generally higher than the conductivity of the cementitious matrix. Moreover, sorption within the crack is expected to be lower than in the matrix and hence cracks in engineered barriers can act as a bypass for radionuclides. Consequently, understanding the effects of crack characteristics on contaminant fluxes from the facility is of utmost importance in a safety assessment. In this paper we numerically studied radionuclide leaching from a crack-containing cementitious containment system. First, the effect of cracks on radionuclide fluxes is assessed for a single repository component which contains a radionuclide source (i.e. conditioned radwaste). These analyses reveal the influence of cracks on radionuclide release from the source. The second set of calculations deals with the safety assessment results for the planned near-surface disposal facility for low-level radioactive waste in Dessel (Belgium); our focus is on the analysis of total system behaviour in regards to release of radionuclide fluxes from the facility. Simulation results are interpreted through a complementary safety indicator (radiotoxicity flux). We discuss the possible consequences from different scenarios of cracks and voids.© 2013 ASME

A single-relaxation-time lattice Boltzmann model for anisotropic advection-diffusion equation based on the diffusion velocity flux formulation

In this paper, we describe a single-relaxation-time (SRT) lattice Boltzmann formulation, which can be effectively applied to anisotropic advection-dispersion equations (AADE). The formulation can be applied to space and time variable anisotropic hydrodynamic dispersion tensor. The approach utilizes diffusion velocity lattice Boltzmann formulation which in the case of AADE can represent anisotropic diagonal and off-diagonal elements of the dispersion matrix by the coupling of advective and diffusive fluxes in equilibrium function. With this approach, AADE can be applied to the SRT lattice Boltzmann formulation using the same equilibrium function and without any changes to collision step nor in the application of boundary conditions. The approach shows good stability even for highly anisotropic dispersion tensor and is tested on selected illustrative examples which demonstrate the accuracy and applicability of the proposed method.

Modelling approach to LILW-SL repository safety evaluation for different waste packing options

The key objective of the work described here was to support the identification of a preferred disposal concept and packaging option for low and short-lived intermediate level waste (LILW-SL). The emphasis of the assessment, conducted on behalf of the Slovenian radioactive waste management agency (ARAO), was the consideration of several waste treatment and packaging options in an attempt to identify optimised containment characteristics that would result in safe disposal, taking into account the cost-benefit of alternative safety measures. Waste streams for which alternative treatment and packaging solutions were developed and evaluated include decommissioning waste and NPP operational wastes, including drums with unconditioned ion exchange resins in overpacked tube type containers (TTCs). For decommissioning wastes, the disposal options under consideration were either direct disposal of loose pieces grouted into a vault or use of high integrity containers (HIC). In relation to operational wastes, three main options were foreseen. The first is overpacking of resin containing TTCs grouted into high integrity containers, the second option is complete treatment with hydration, neutralization, and cementation of the dry resins into drums grouted into high integrity containers and the third is direct disposal of TTCs into high integrity containers without additional treatment. The long-term safety of radioactive waste repositories is usually demonstrated with the support of a safety assessment. This normally includes modelling of radionuclide release from a multi-barrier near-surface or deep repository to the geosphere and biosphere. For the current work, performance assessment models were developed for each combination of siting option, repository design and waste packaging option. Modelling of releases from the engineered containment system (the ‘near-field’) was undertaken using the AMBER code [1]. Detailed unsaturated water flow modelling was undertaken using the HYDRUS code [2], where the degree of engineered barrier degradation with time is accounted for in each packaging option. Water fluxes relating to each degradation level were then incorporated into the AMBER models for further radionuclide transport calculations appropriate to each packing solution. The approach proved to be highly flexible, transparent and effective in terms of calculation time. Results demonstrate that all waste streams could be accepted at the preferred site with the surface repository option, under the condition that all decommissioning waste would be grouted into high integrity containers. The use of high integrity containers is also recommended for all other waste streams. Results from the detailed analysis further showed that in-drum-dried ion exchange resins in TTCs would be acceptable when grouted into high integrity containers, thereby avoiding the need for complicated processing and repackaging.Copyright