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

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.

Study on combustible gas generation and radionuclide release during underwater handling of the AM reactor spent fuel

Introduction. The paper addresses studies on the accumulation of combustible gases during underwater handling simulations for the leaky spent nuclear fuel from the AM reactor. Two fuel compositions were studied- uranium-molybdenum dispersed in magnesium and uranium carbide dispersed in calcium. Methods. The 137Cs release rate was measured during underwater storage of the uranium-molybdenum fuel. The kinetics of hydrogen release for both fuels and methane release for the carbide SNF were obtained. The kinetics approximate most with exponential dependences that formally correspond to first-order chemical reactions. A contribution of radiolytic hydrogen to the gases generated during the experiments was estimated. It was demonstrated that the determining source of the gases is the chemical interaction between the spent fuel and the water. The experiment with the uranium-molybdenum fuel demonstrated a pronounced passivation effect of the chemical processes on the fuel surface due to insoluble corrosion products. For the carbide SNF, an incubation period of about 20 hours was observed followed by an intensive release of hydrogen and methane. Results. The obtained results were subject to a comparative analysis against publications on the behavior of the fuel components in water. Conclusion. The findings can be applied to justify fire and explosion safety of underwater handling techniques for the damaged spent nuclear fuel with the considered fuel compositions (the spent fuel from reactors AM, AMB, EGP-6, etc.), e.g., to justify underwater preparations of the AMB spent fuel for reprocessing.

An Alternative Conceptual Model for the Spent Nuclear Fuel–Water Interaction in Deep Geologic Disposal Conditions

For the long-term safety assessment of direct disposal of spent nuclear fuel in deep geologic repositories, knowledge on the radionuclide release rate from the UO2 matrix is essential. This work provides a conceptual model to explain the results of leaching experiments involving used nuclear fuel or simulant materials in confirmed reducing conditions. Key elements of this model are: direct effect of radiation from radiolytic species (including defects and excited states) in the solid and in the first water layers in contact with its surface; and excess H2 may be produced due to processes occurring at the surface of the spent fuel and in confined water volumes, which may also play a role in keeping the spent fuel surface in a reduced state. The implication is that the fractional radionuclide release rate used in most long-term safety assessments (10−7 year−1) is over estimated because it assumes that there is net UO2 oxidation caused by radiolysis, in contrast with the alternative conceptual model presented here. Furthermore, conventional water radiolysis models and radiation chemical yields published in the literature are not directly applicable to a heterogeneous system such as the spent fuel–water interface. Suggestions are provided for future work to develop more reliable models for the long-term safety assessment of spent nuclear fuel disposal.

Calculation of the Dose for Public Individuals Due to a Severe Accident at the Angra 2 Nuclear Plant, Brazil

Through a severe accident at nuclear power plant Angra 2, the whole body dose effective of the individuals members of the public located in the Emergency Planning Zones (EPZs) will be calculated, and later, the protective actions in these EPZs will be analyzed. Two different scenarios of radionuclide release into the atmosphere will be considered. In the first scenario, 2 h of the release of Xe, Cs, Ba, and Te, and the second scenario, 168 h of release.

Scenarios and Pathways of Radionuclide Releases from Near-Surface Waste Disposal Facilities: A Brief Overview of Historical Evidence

The very low-level waste (VLLW) produced during decommissioning of nuclear facilities can be suitable for disposal in landfill type facilities. Considering the similarities in design, the experience gained in near-surface disposal of radioactive waste in trenches and vaults is relevant to the issue of VLLW disposal in landfills. This paper presents a brief review of internationally reported cases of radionuclide releases from near-surface disposal facilities. Based on this review, the conclusions are made that the following radionuclide release and exposure scenarios should be accounted for in safety assessment of VLLW disposal in landfills: i) leaching from waste to groundwater by atmospheric precipitations; ii) bath-tubing scenario; iii) scenarios caused by extreme meteorological and hydrological events (erosion, flooding, etc.); iv) human intrusion. The gaseous transport deserves attention for a number of relevant radionuclides, such as (C-14, Rn-222, etc.). In addition, the possibility of early degradation of engineered containment structures (soil covers, bottom seals) should be cautiously considered.