Guido Deissmann

New insights into phosphate based materials for the immobilisation of actinides

Abstract This paper focuses on major phosphate-based ceramic materials relevant for the immobilisation of Pu, minor actinides, fission and activation products. Key points addressed include the recent progress regarding synthesis methods, the formation of solid solutions by structural incorporation of actinides or their non-radioactive surrogates and waste form fabrication by advanced sintering techniques. Particular attention is paid to the properties that govern the long-term stability of the waste forms under conditions relevant to geological disposal. The paper highlights the benefits gained from synergies of state-of-the-art experimental approaches and advanced atomistic modeling tools for addressing properties and stability of f-element-bearing phosphate materials. In conclusion, this article provides a perspective on the recent advancements in the understanding of phosphate based ceramics and their properties with respect to their application as nuclear waste forms.

Implications of Grain-Scale Mineralogical Heterogeneity for Radionuclide Transport in Fractured Media

The geological disposal of nuclear waste is based on the multi-barrier concept, comprising various engineered and natural barriers, to confine the radioactive waste and isolate it from the biosphere. Some of the planned repositories for high-level nuclear waste will be hosted in fractured crystalline rock formations. The potential of these formations to act as natural transport barriers is related to two coupled processes: diffusion into the rock matrix and sorption onto the mineral surfaces available in the rock matrix. Different in situ and laboratory experiments have pointed out the ubiquitous heterogeneous nature of the rock matrix: mineral surfaces and pore space are distributed in complex microstructures and their distribution is far from being homogeneous (as typically assumed by Darcy-scale coarse reactive transport models). In this work, we use a synthetically generated fracture–matrix system to assess the implications of grain-scale physical and mineralogical heterogeneity on cesium transport and retention. The resulting grain-scale reactive transport model is solved using high-performance computing technologies, and the results are compared with those derived from two alternative models, denoted as upscaled models, where mineral abundance is averaged over the matrix volume. In the grain-scale model, the penetration of cesium into the matrix is faster and the penetration front is uneven and finger-shaped. The analysis of the cesium breakthrough curves computed at two different points in the fracture shows that the upscaled models provide later first-arrival time estimates compared to the grain-scale model. The breakthrough curves computed with the three models converge at late times. These results suggest that spatially averaged upscaled parameters of sorption site distribution can be used to predict the late-time behavior of breakthrough curves but could be inadequate to simulate the early behavior.

Continuum-based DFN-consistent numerical framework for the simulation of oxygen infiltration into fractured crystalline rocks

We present an enhanced continuum-based approach for the modelling of groundwater flow coupled with reactive transport in crystalline fractured rocks. In the proposed formulation, flow, transport and geochemical parameters are represented onto a numerical grid using Discrete Fracture Network (DFN) derived parameters. The geochemical reactions are further constrained by field observations of mineral distribution. To illustrate how the approach can be used to include physical and geochemical complexities into reactive transport calculations, we have analysed the potential ingress of oxygenated glacial-meltwater in a heterogeneous fractured rock using the Forsmark site (Sweden) as an example. The results of high-performance reactive transport calculations show that, after a quick oxygen penetration, steady state conditions are attained where abiotic reactions (i.e. the dissolution of chlorite and the homogeneous oxidation of aqueous iron(II) ions) counterbalance advective oxygen fluxes. The results show that most of the chlorite becomes depleted in the highly conductive deformation zones where higher mineral surface areas are available for reactions.

Microtomography-based Inter-Granular Network for the simulation of radionuclide diffusion and sorption in a granitic rock

Field investigation studies, conducted in the context of safety analyses of deep geological repositories for nuclear waste, have pointed out that in fractured crystalline rocks sorbing radionuclides can diffuse surprisingly long distances deep into the intact rock matrix; i.e. much longer distances than those predicted by reactive transport models based on a homogeneous description of the properties of the rock matrix. Here, we focus on cesium diffusion and use detailed micro characterisation data, based on micro computed tomography, along with a grain-scale Inter-Granular Network model, to offer a plausible explanation for the anomalously long cesium penetration profiles observed in these in-situ experiments. The sparse distribution of chemically reactive grains (i.e. grains belonging to sorbing mineral phases) is shown to have a strong control on the diffusive patterns of sorbing radionuclides. The computed penetration profiles of cesium agree well with an analytical model based on two parallel diffusive pathways. This agreement, along with visual inspection of the spatial distribution of cesium concentration, indicates that for sorbing radionuclides the medium indeed behaves as a composite system, with most of the mass being retained close to the injection boundary and a non-negligible part diffusing faster along preferential diffusive pathways.

Simulating Oxygen Intrusion into Highly Heterogeneous Fractured Media Using High Performance Computing

Fractured crystalline rocks are under consideration by several countries as host formations for high-level nuclear waste repositories. The redox evolution in these host rock formations is an important issue for the stability and safety of these disposal sites. If, for example, during a glaciation/deglaciation event, oxygen-rich glacial meltwater penetrates to the depth of the planned repository, some of the engineered barriers would be adversely affected. Moreover, oxidizing conditions would increase the solubility and mobility of many radionuclides. Reactive transport simulations, which are typically used to assess the redox buffering capacity of these host rock formations, are computationally demanding, and thus, calculations for the evaluation of oxygen penetration are usually carried out over simplified geometries and the heterogeneity of the site, both physical (e.g., variability in the groundwater flow field and the kinematic porosity) and mineralogical (e.g., variability in the abundance of Fe(II)-bearing minerals), is usually represented in a simplified fashion. Here, it is shown how a recently developed numerical framework, combined with high performance computing technologies, allows the full geometrical, physical and mineralogical complexity of the site under study to be efficiently included in these types of reactive transport calculations. A synthetically generated realistic three-dimensional fractured medium is used to assess oxygen penetration patterns and their dependence on both the hydrogeological conditions and the availability of Fe(II)-bearing minerals. The results of the calculations point out the significant influence of both physical and mineralogical heterogeneity on the oxygen penetration patterns, thus highlighting the importance of a model parameterisation consistent with the site complexity.

Integration of Life Cycle Assessments in the decision-making process for environmental protection measures and remedial action at active and abandoned mining sites

Life-Cycle Assessments (LCA) have increasingly been used to evaluate the environmental performance of industrial processes and products and have been employed as a basis in decision making processes in the public and industrial sector. In contrast, assessment methodologies for environmental protection measures and remediation measures with respect to their consumption of natural resources are generally poorly developed. In this paper a methodology for LCA of environmental protection measures and remedial actions is suggested, using the remediation/reclamation of mining sites as an example.

Development and application of knowledge-based source-term models for radionuclide mobilisation from contaminated concrete

Concrete materials in nuclear facilities may become activated or contaminated by various radionuclides through different mechanisms. Consequently, decommissioning and dismantling of these facilities produce considerable quantities of these materials (e.g. concrete structures, rubble), which are at least potentially contaminated with radionuclides and which must be managed safely and cost-effectively. In this paper, we present results from a research project that aims at the development of source-term models for the mobilization of radionuclides from contaminated concrete. The objective of this task was to clarify whether a more realistic sourceterm description could be beneficial for optimization of the management of decommissioning wastes by reducing the amount of material for disposal as radioactive waste as well as by saving natural resources due to the recycling of building materials. To identify important parameters and processes that affect the release rates of radionuclides, we evaluated the chemical behavior and the solid speciation of radionuclides in concrete materials and the influence of factors like concrete properties, source/pathway of contamination, and the scenario-specific chemical environment and hydraulic regime. Furthermore, concrete degradation processes and their influence on contaminant mobilization were addressed. On this basis, source-term models were developed to describe the radionuclide release by (i) the dissolution of radionuclide containing solid phases, (ii) the desorption of radionuclides from surfaces, and/or (iii) the leaching of radionuclides from a solid matrix without disrupting its structure. These source-term models were parameterized for probabilistic simulations of various release options, including the reuse of recycled building materials, the disposal of rubble in inert and municipal landfills as well as the on-site disposal of concrete materials (e.g. foundations remaining in the ground, in situ burial of rubble). For some scenarios and radionuclides, the calculated release rates were between one and two orders of magnitude lower than those used in former generic calculations. Based on the results of stochastic simulations, the consequences of the use of a more realistic source-term for dose assessments with respect to clearance/recycling of contaminated concrete will be illustrated and discussed.

Radioactive Contamination of Concrete: Uptake and Release of Radionuclides

Concrete in nuclear installations may become contaminated by various radionuclides. Consequently, decommissioning and dismantling produce considerable quantities of potentially contaminated materials that must be managed safely and cost-effectively. In this paper we present preliminary results from a research project that aims to improve knowledge about release behaviour of radionuclides from contaminated concrete, and that proposes a scientific approach to calculating the source term for radiological dose assessment for the various management options (e.g. direct reuse, recycling, disposal of rubble). The first step is to consider which nuclides are likely to have contaminated concrete, where they might be located in concrete, and the extent to which they are chemically bound to concrete constituents. Relevant radionuclides include 60 Co, 63 Ni, 90 Sr, 137 Cs, 129 I, U, Pu, Am and other actinide elements. Some nuclides are likely to be bound in specific solid phases and others are sorbed to greater or lesser degrees. The proposed modelling of releases from concrete takes into account the chemical behaviour (speciation, sorption and solubility) of the individual radionuclide contaminants and their binding to concrete phases. Other important factors that will influence release are mechanical and chemical condition of concrete, including cracking, carbonation, sulfate attack and degree of water saturation. Model calculations illustrate the potential release processes of desorption-diffusion, leaching (shrinking core model) and dissolution of discrete solid phases. For example, a scoping calculation suggests that 50-year old concrete may be contaminated with 129 I to about 1 cm depth from the surface or more if the concrete is degraded, and that subsequent release will occur slowly by diffusion. Strongly sorbed or particulate nuclides such as Pu are likely to remain at the surface. Predicting the behaviour of some nuclides (e.g. Ni, U) is more uncertain because of uncertainty in the key parameters and their dependence on the local chemical conditions. Release models and source terms have been developed as the starting point for (i) the modelling of radiological consequences (i.e. dose assessments) of disposal options for building materials from nuclear installations and the optimisation of the disposal process (i.e. selection of cost-effective and reasonable disposal options), and (ii) the assessment of recycling/reuse options of slightly contaminated materials in order to reduce the amount of waste for disposal.Copyright

Assessment of the release behaviour of 14C from irradiated nuclear graphite from a German research reactor

The behaviour of irradiated graphite (i-graphite) from a German research reactor was investigated under different boundary conditions for assessment of 14C-speciation and the release kinetics into the aqueous and gas phase. The results showed that most of released 14C remains in the solution. Under near neutral conditions, volatile 14C is predominantly in form of CO2 (i.e. 90%) with some minor fraction of 14CO + 14Corg. Almost negligible 14C release into the gas phase was detected under cementitious conditions, although the total 14C release increases. Evaluated release rates of 14C are discussed in the context of i-graphite disposal in the GDF Schacht Konrad.

Release behaviour of radionuclides from contaminated concrete materials

During the decommissioning and dismantling of nuclear facilities large quantities of radioactively contaminated concrete materials arise, which must be managed safely and cost-effectively. This paper provides an overview of a research project dealing with the development of source terms for the mobilisation of radionuclides from contaminated concrete. Key parameters to be taken into account comprise contamination sources/pathways, the physical, chemical, and min-eralogical properties of the concrete and their alteration with time, as well as the environmental conditions depending on the chosen disposal/reuse option.