Bentonites are considered to be the most suitable materials for the multibarrier system of high-level radioactive waste repositories. Since BCV bentonite has been proved to be an ideal representative of Czech Ca-Mg bentonites in this respect, it has been included in the Czech Radioactive Waste Repository Authority (SÚRAO) buffer and backfill R&D programme. Detailed knowledge of processes in the material induced by thermal loading provides invaluable assistance regarding the evolution of the material under repository conditions. Samples of both original BCV 2017 bentonite and the same material thermally treated at 200 °C were characterised by means of chemical analysis, powder X-ray diffraction, infrared spectroscopy, thermal analysis, cation exchange capacity, specific surface area (BET) measurements, the determination of the swell index, the liquid limit, the swelling pressure and water retention curves. The smectite in BCV 2017 bentonite comprises Ca-Mg montmorillonite with a significant degree of Fe³⁺ substitution in the octahedral sheet. Two main transformation processes were observed following heating at 200 °C over 27 months, the first of which comprised the dehydration of the montmorillonite and the subsequent reduction of the 001 basal distance from 14.5 Å (the original BCV 2017) to 9.8 Å, thus indicating the absence of water molecules in the interlayer space. The second concerned the dehydration and partial dehydroxylation of goethite. With the exception of the dehydration of the interlayer space, the PXRD and FTIR study revealed the crystallochemical stability of the montmorillonite in BCV 2017 bentonite under the selected experimental conditions. The geotechnical tests indicated no major changes in the mechanical properties of the thermally treated BCV 2017 bentonite, as demonstrated by the similar swelling pressure values. However, the variation in the swell index and the gradual increase in the liquid limit with the wetting time indicated a lower hydration rate. The retention curves consistently showed the lower retention capacity of the thermally treated samples, thus indicating the incomplete re-hydration of the thermally treated BCV 2017 exposed to air humidity and the difference in its behaviour compared to the material exposed to liquid water.

The Josef Mock-Up experiment, conducted in the Josef Underground Laboratory, simulated the vertical disposal of a spent nuclear fuel container surrounded by compacted, artificially Na₂CO₃-activated Czech bentonite (“Bentonite 75”) under repository-relevant conditions. This approx. 50% scale model (height 2.2 m, diameter 0.7 m) was managed by the Radioactive Waste Repository Authority and constructed and operated by the Czech Technical University in Prague. After nearly a decade of thermal loading at ~ 90 °C, the system was dismantled in late 2022 and extensively sampled. Over 220 bentonite samples from 18 layers were analysed to assess physical, geochemical, and microbiological changes. The distribution of dry density (between 1350 and 1570 kg·m⁻³) and degree of saturation (Sr; 0.80–1.05) revealed partial homogenisation, with the highest saturation in upper layers due to water ingress. Geochemical/mineralogical analyses showed changes in smectite basal diffraction line (from 12.57 to 11.96–9.98 Å), composition of exchangeable cations (Na⁺ depletion, Ca²⁺/Mg²⁺ enrichment) and composition of water leachates (Na⁺ depletion). The elevated siderite content was analysed near corrosion samples and other metal construction components. These changes were attributed to the combined effects of thermal loading and the interaction of water. Microbiological analysis confirmed near-sterility of the bentonite, with minimal bacterial survival and poor colonisation potential of the input material, likely due to prior Na₂CO₃ activation. The results provided data for updating thermo-hydro-mechanical and geochemical models of bentonite barrier evolution and information on the stability of the bentonite barrier in realistic conditions.