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We explored the use of selected layered double hydroxides (LDHs) of different compositions and obtained by means of different routes (i.e., coprecipitation, anionic exchange and reconstruction) as iodine/iodate adsorbents. The influence of speciation (iodide vs. iodate) on iodine uptake was rather strong, resulting in much lower iodide incorporation. The Fourier transform of iodine K X-ray absorption edge data (EXAFS) of all iodate-LDHs showed a single I-O scattering path of 1.8 Å . Thermal stability and leaching experiments showed that the incorporated iodate and iodide were rather loosely bound in the interlayer space and were easily released under heating to 180°C and leaching with Milli-Q water and brine solution.

To reduce uncertainties in determining the source term and evolving condition of spent nuclear fuel is fundamental to the safety assessment. ß-emitting nuclides pose a challenging task for reliable, quantitative determination because both radiometric and mass spectrometric methodologies require prior chemical purification for the removal of interfering activity and isobars, respectively. A method for the determination of 90Sr at trace levels in nuclear spent fuel leachate samples without sophisticated and time-consuming procedures has been established. The analytical approach uses a commercially available automated pre-concentration device (SeaFAST) coupled to an ICP-DRC-MS. The method shows good performances with regard to reproducibility, precision, and LOD reducing the total time of analysis for each sample to 12.5 min. The comparison between the developed method and the classical radiochemical method shows a good agreement when taking into account the associated uncertainties.

Abstract. Safety concepts regarding nuclear waste disposal in underground repositories generally rely on a combination of engineered and geological barriers, which minimize the potential release of radionuclides from the containment-providing rock zone or even their transport into the biosphere. Cementitious materials are used for conditioning of certain nuclear waste types, as components of waste containers and overpacks, as well as being constituents of structural materials at the interface between backfilling and host rock in some repository concepts. For instance, the preferred option for the disposal of high-level waste (HLW) in Belgium is based on the supercontainer design, which consists of a carbon steel overpack surrounded by a thick concrete buffer (Bel et al., 2006). In the event of formation water interacting with cementitious materials, pore water solutions characterized by (highly) alkaline pH conditions will form. These boundary conditions define the chemical response of the radionuclides, but also influence the behaviour of neighbouring components of the multi-barrier system, e.g. bentonitic or argillaceous backfilling and host rock. Hardened cement paste or Sorel cement are considered to be main sorbing materials present in the near field of repositories for low- and intermediate-level waste (L/ILW). Hence, interactions of radionuclides with cementitious materials represent a very important mechanism retarding their mobility and potential migration from the near field (Wieland, 2014; Ochs et al., 2016). While the quantitative description of the sorption processes (usually in terms of sorption coefficients, i.e. Kd values) is a key input in the safety analysis of nuclear waste repositories, detailed mechanistic analysis and understanding of sorption phenomena provide additional scientific arguments and important process understanding, and thus enhance both the quality of safety arguments and the overall confidence in the safety assessment process. Research at KIT-INE dedicated to the interaction of cementitious materials with radionuclides is conducted in the context of different repository concepts, including clay (low- and high-ionic strength conditions), crystalline rock or rock salt. Experimental and theoretical studies are performed within the framework of national (GRAZ, BMWi) and international (CEBAMA and EURAD-CORI, EU Horizon 2020 Programme) projects, extending to third-party projects with several waste management organizations in Europe, e.g. SKB (Sweden), ONDRAF-NIRAS (Belgium) or BGE (Germany). The combination of classical experimental (wet chemistry) methods, advanced spectroscopic techniques and theoretical calculations provides both an accurate quantitative evaluation and a fundamental understanding of the sorption processes. Examples of recent studies at KIT-INE on radionuclide behaviour in cementitious systems in the context of both L/ILW and HLW will be presented in this contribution to explain methodologies, scientific approaches and results. The present state of knowledge as well as main remaining uncertainties affecting the retention processes of radionuclides in cementitious environments under different conditions will be critically discussed, also in view of current international research activities and repository 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.

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