In the framework of interim storage of spent MOX fuel assemblies in pools during several decades, it is necessary to take into account an incidental scenario. In the event of cladding failure, corrosion processes can lead to a deterioration/damage of the failed rod and to a radionuclide release into water. In order to study spent MIMAS MOX fuel behaviour in pools, a leaching experiment with chemical and radiological conditions as realistic as possible have been performed. In a MOX fuel rod irradiated up to 47GWj/t, a cladded segment was cut and polished on one side (opened section). It was immersed in a boric acid solution (2g/l), at 27°C, under air and under ⁶⁰Co-source, providing a gamma-irradiation field of about 70Gy/h during the whole leaching time (4 months). Solution has been regularly sampled and analysed in order to measure radionuclides releases and the concentration of hydrogen peroxide (H₂O₂) formed by water radiolysis. The spent fuel segment was regularly taken out from the solution to characterize its surface evolution (by optical microscopy coupled with Raman spectroscopy).

Results showed that uranium releases stabilized after 3 months (total released fraction of about 1.6x10^-5 of the inventory), and the final analysis of the leachate showed an important solid fraction (>0.45µm) containing uranium. Solution analyses, surface observations and Raman spectra acquired on it enabled to follow the massive precipitation of studtite, an uranium peroxide formed through the reaction of UO₂²⁺ with the H₂O₂ generated by the radiolysis of water. The studtite precipitation seems to control the uranium concentrations in solution for these conditions. Uranium is not a good tracer of spent fuel matrix alteration, conversely to ⁹⁰Sr and ¹³⁷Cs which are linearly released in the solution during the leaching. These experimental results are compared to the modelling data obtained from radiolysis and solution chemistry simulations.

The disposal of spent nuclear fuels (SNF) in a deep geological repository (DGR) is regarded as the best practical waste management option in many countries. The long-term safety of a DGR over an assessment time frame of up to one million years necessitates a comprehensive understanding of the corrosion behaviour of SNF once the waste canister is breached and groundwater comes into contact. Although various studies have addressed this topic in the last decades, some of the processes contributing to the (radiolytic) matrix corrosion of SNF in the generally reducing repository environment are not fully understood. Furthermore, only limited efforts were deployed to study the corrosion behaviour of irradiated MOX fuels. To examine the effects of environmental conditions on SNF corrosion, the SF-ALE project (Spent Fuel Autoclave Leaching Experiments) was started as a collaboration between the Belgian Nuclear Research Centre SCK CEN and the Forschungszentrum Jülich GmbH.

Within SF-ALE, irradiated MOX fuel rod segments (burn-up between 29 GWd/t_HM and 52 GWd/t_HM) were leached in bicarbonate water at neutral pH and in synthetic cementitious water at pH 13.5 under reducing atmosphere (4 vol% H₂ in Ar at 40 bars pressure) in order to assess the release of various radionuclides and fission gases over a timeframe of 3.5 years. Following the leaching phase, a post-leaching characterisation of the fuel rod segments was initiated. Scanning electron microscopy analyses revealed that the structure of the MOX fuel matrix was affected differently by the exposure to the varied environmental conditions. Furthermore, secondary phases and alteration products were observed on the fuel surfaces. This contribution introduces initial results of the post-leaching characterisation and their implications to the general understanding of the corrosion behaviour of SNF under repository relevant conditions.

For the safety assessment of a future deep repository for spent nuclear fuel, the rate and mechanism for dissolution of fission products and actinides from the fuel is a key parameter. For the vast majority of the fuel rods, the cladding is expected to be intact at the time of disposal. There is however a small fraction of rods where the cladding has failed and the fuel has been exposed to water and/or air prior to encapsulation. In these rods an alteration of the fuel matrix can be expected. In order to predict the mechanism and dissolution rate of radionuclides from failed fuel, the effects of this matrix alteration need to be investigated. Although the leaching behaviour of spent nuclear fuel under deep repository conditions has been studied extensively, studies of failed fuel under these conditions are relatively scarce. However, at the Studsvik Hot Cell laboratory, aerated leaching tests of spent fuel started in the 1980s and two of these were finalized only recently. This has provided an opportunity to investigate the effects of long-term contact with water under aerated conditions similar to wet interim storage conditions.

As reported previously, when finalizing the leaching studies after 37 years leaching time, the fuel samples were subject to visual inspection, X-ray diffraction investigations and leaching in carbonate solution in order to study the formation and properties of secondary phases (mainly studtite and metashoepite). In the present work we have performed scanning electron microscopy including WDS-analysis as well as Laser Ablation-ICP-MS studies on one of the fuel samples. The results show that the secondary phases formed upon long term exposure to aerated deionized water are depleted of most radionuclides, with exception of Cs, and to some extent Rb and Sr, which seem to be retained in the secondary phase.

During interim storage of fuel assemblies in pools the prospect of a through-wall cladding defect must be taken into consideration. The pool water is subjected to a strong radiation field producing oxidizing species, the main molecular species of which is hydrogen peroxide. When brought into contact with the fuel, hydrogen peroxide generated by water radiolysis leads to the precipitation of studtite (UO₂)(O₂)(H₂O)₂ at the surface of the fuel. This phase having a density lower than that of the fuel can induce a worsening of the defect and thus impact the mechanical strength of the rods. It is therefore important to understand their formation mechanisms at the surface of the spent fuel.

In order to assess this problem, an experimental approach aiming to understand its formation mechanism in the presence of alpha radiolysis of water was conducted. Thus, an ¹⁸O-labeled solution was irradiated with He²⁺ ions through a UOx target, itself leached by the radiolysed solution. An in-situ Raman monitoring of the studtite precipitation kinetics was carried out. The use of ¹⁸O-labeled water was chosen, in order to distinguish the air oxygen dissolved in the solution from that contained in the water.

The absence of the presence of ¹⁶O in the peroxide bond added to the results of the Chemsimul software, reveals that contrary to gamma radiolysis, the oxygen dissolved in the water does not intervene in the mechanism of production of studtite in alpha radiolysis. We deduce that the process of H₂O₂ formation is the result of the primary radiolytic yield during the heterogeneous chemistry step by the OH°+ OH° recombination. This work teaches us that the mechanisms of H₂O₂ production in aerated water, leading to the peroxide bonds of studtites, are dependent on TEL effects.

Experimental research on active materials often goes along with the generation of compositionally complex waste streams for which a suitable route towards safe (interim) storage is lacking. The complexity of the streams invokes the need for tailored solutions for the individual components. Research on possible reuse but mainly final disposal of the spent fuel is an important part of the new nuclear fuel concepts. A prerequisite for any route is that the waste form can be accepted by the national organizations for waste disposal.

At NRG, research on molten salt reactor (MSR) fuels, both fluorides and chlorides, is ongoing. A part of the research is dedicated to waste handling following irradiation experiments in the HFR Petten. The created spent fuel waste will be finally disposed by the national organization for waste disposal (COVRA). To get this new waste accepted, it must be first fully characterized and the fluoride and chloride waste must be transformed to chemically stable and acceptable waste streams. It is foreseen that if new forms of waste are offered for disposal, additional tests related to final disposal are required; the chemical stability of the immobilized waste forms, most notably cemented waste, must be tested by specific leaching experiments to meet the waste acceptance criteria.

A comprehensive literature survey has been done to summarize the possible ways of handling the MSR waste including molten salt specific challenges (such as radiolysis leading to halide gas formation). Based on the review, different routes have been identified, and have been applied to experimental cold and ‘semi-hot’ tests. Two of the tested methods show the most promising results; i) direct defluorination/dechlorination method and ii) vitrification using ironphosphate glass.

The proposed and tested route for waste handling will be finally applied at NRG to irradiated and fully characterized MSR fuels from the NRG MSR irradiation programme.