Conference Agenda

In the framework of radioactive waste management from nuclear power plants, research and medical applications, the most difficult challenge is the identification of the best waste route suitable for predisposal. From the point of view of immobilization of radioactive waste, several techniques, depending on the waste (stream) activity, such as vitrification, bitumen and in particular, cementation, have been adopted. Cementation in ordinary Portland cement or blended matrices has been demonstrated to be effective in obtaining homogeneous waste forms on cost-effective basis. An alternative to cementitious waste form is alkali-activated materials, which combine better chemical durability with lower environmental impact. In recent years, R&D efforts have been concentrated on investigating their possible use for direct waste conditioning.

This study focuses on identifying a new immobilization matrix for Pb and Pb/Bi alloys that are used as coolants in lead-cooled small-modular and fast reactors respectively. The base material for those matrices is Fe-rich glasses (Vitrified Bauxite Residue), which is used as a precursor for an alkali-activated material. After having described the chemical-physical characteristics of the matrix, the study attempts to find an appropriate formulation for the realization of the alkali-activated material, which is able to comply with the waste acceptance criteria (e.g. strength) and has good fresh properties. Specifically, a key aspect of this research concerns the mineralogical and microstructural characterization of the waste form, with the aim of highlighting the binding mechanism of Pb and Bi within the alkali-activate matrix.

Results show that good compatibility can be obtained with Pb but not with Pb/Bi, since Bi introduces variability in terms of polarizability, reactivity and compatibility with geopolymer network. The alloy can form intermetallic compounds that interact weakly with the matrix, causing instability, component migration or the formation of unwanted phases that compromise the strength of the material. The presence of Bi also favors the formation of crystalline phases or microcracks, reducing their overall compatibility with the system. Good mechanical and fresh properties of the alkali-activated material itself are expected, since the alkali-activated material, like cement, can reach a good mechanical strength (compatible with the Italian criteria, requiring at last 10 MPa compressive strength). With this research we demonstrate that alkali-activated materials are well suited to replace a large part of traditional cement.

Most future sodium-cooled fast neutron reactors will operate uranium−plutonium mixed oxide (MOX) fuels, which may also contain minor actinides (MAs) for transmutation purposes. Pu/Metal and MA/Metal ratios will likely be set at 25~35 mol% and 1~5 mol%, respectively.

Around the world, MOX is produced industrially by powder metallurgy and reactive sintering to obtain dense pellets with appropriate cation distribution and Oxygen/Metal (O/M) ratio. Oxygen stoichiometry is a key characteristic that must be carefully controlled as it impacts most of the fuel’s thermo-physicochemical properties [1]. Over the last six decades, extensive research has been carried out to measure and control the O/M ratio of MOX, underlining the importance of this parameter in developing these nuclear fuels.

The oxygen stoichiometry of MOX varies during sintering due to solid/gas exchanges, but controlling these variations is difficult as the kinetics of non-equilibrium reactions induces them. To date, industrial sintering of MOX has been carried out in a constant atmosphere for technological reasons, generally using a mixture of Ar + 4~10% H₂ with or without added moisture. However, recent studies have shown that adapting the oxygen partial pressure (pO₂) along the thermal profile holds promise for improving the characteristics of the sintered fuel pellets [2].

The present study focuses on measuring the oxygen chemical diffusion at elevated temperatures (1773, 1873, and 1923 K) using TGA in U_0.698Pu_0.289Am_0.013O_2−x, a composition assumed to be representative of future SFR fuels. A new experimental strategy has been developed to study the chemical diffusion of oxygen in MOX as a function of the type of lattice defect responsible for the deviation from stoichiometry.

The experimental data collected during this study and their analysis have shown that:

・The chemical diffusion of oxygen depends on the O/M ratio of the MOX pellets.

・With increasing hypo-stoichiometry, reducing MOX becomes increasingly difficult (i.e. it takes more and more time).

・The variations in the reduction time versus the O/M ratio change range follow a polynomial trend. As an example, the time needed to reduce MOX from O/M = 1.96 to 1.95 takes ×2.5 more time than reducing MOX from O/M = 2.00 to 1.99.

The experimental data collected during this study and their analysis can be used to ingeniously tailor MOX sintering conditions for fast reactors.

[1] C. Duriez, J.-P. Alessandri, T. Gervais, Y. Philipponneau, Thermal conductivity of hypostoichiometric low Pu content (U,Pu)O_2−x mixed oxide, J. Nucl. Mater. 277 (2000) 143–158. https://doi.org/10.1016/S0022-3115(99)00205-6.

[2] S. Nakamichi, S. Hirooka, M. Kato, T. Sunaoshi, A.T. Nelson, K.J. McClellan, Effect of O/M ratio on sintering behavior of (Pu_0.3U_0.7)O_2−x, J. Nucl. Mater. 535 (2020) 152188. https://doi.org/10.1016/j.jnucmat.2020.152188.

Over the past years, the Slovenian Agency for Radioactive Waste Management (ARAO) has made significant advancements in radioactive waste management. This paper presents key developments, challenges, and achievements in four major areas: the planning and construction of the low- and intermediate-level radioactive waste (LILW) disposal facility, the management of institutional radioactive waste (IRAW), the planning of high-level radioactive waste (HLW) and spent fuel (SF) disposal, and the long-term surveillance and maintenance of closed disposal sites for mining and hydrometallurgical tailings from the former Žirovski vrh uranium mine.

The construction of the LILW disposal facility in Vrbina, Krško, has progressed significantly, with the acquisition of a construction permit in 2023. The project follows internationally recognized safety standards, ensuring the repository's operational and long-term safety and sustainability. Construction works have commenced in mid 2023, marking a crucial step towards the finalization of Slovenia’s first dedicated repository for LILW.

ARAO has successfully executed several shipments of disused sealed radioactive sources abroad for recycling, contributing to the reduction of domestic radioactive waste storage requirements. Furthermore, a peer review mission was conducted, confirming ARAO’s compliance with international standards and best practices in radioactive waste management.

Significant efforts have been made in planning HLW and SF disposal solutions. Technical studies have been conducted to assess potential solutions for long-term disposal, with a focus on deep geological repository (DGR) site selection, safety analysis, disposal technologies, construction, operation, decommissioning and closure.

ARAO has established a long-term monitoring system for the closed uranium mining waste sites at Žirovski vrh. Continuous environmental monitoring, including radiation measurements in air, water, and soil, ensures the safety of the surrounding environment and local communities.

Despite notable progress, challenges remain. Future efforts will focus on completing the LILW repository, advancing HLW and SF disposal plans, and enhancing long-term waste management sustainability.

This paper highlights ARAO’s achievements and ongoing initiatives in radioactive waste management. The agency remains committed to implementing safe, transparent, and efficient waste management practices, aligning with international standards and best practices.

Keywords: radioactive waste management, LILW repository, institutional radioactive waste, HLW, SF, long-term monitoring, ARAO, Slovenia