The dissolution of the United Kingdom’s vitrified high-level-waste simulant, CaZn MW28, was investigated following the Product Consistency Test-B protocol for 112 d at 90 °C and in ultra-high-quality water. Residual rate dissolution (stage II) and rate resumption (stage III), after 28 d, was observed. Thermodynamic modelling suggested that solutions were saturated with respect to Mg- and Zn-bearing phases, and the presence of Mg- and Zn-smectite clays was tentatively observed. The formation of these phases was concurrent with a significant increase in the dissolution rate, similar to Stage III behavior seen in other nuclear waste simulant glass materials, indicating that the addition of Mg and Zn to high-level-waste glass (7.3 wt. % combined) significantly influences the dissolution rate.

The three generically accepted stages of glass corrosion are reviewed with focus on final stage termed alteration rate renewal (or resumption) stage when the glass may re-start corroding with the rate similar to that at the initial stage. It is emphasized that physical state and physical changes that occur in the near-surface layers can readily lead to an effective increase of leaching rate which is similar to alteration rate renewals. Experimental data on long-term (during few decades) corrosion of radioactive borosilicate glass K26 designed to immobilize high-sodium operational NPP radioactive waste evidence on resumption-like effects of radionuclides (137,134Cs) leaching. The cause of that was however related not to chemical changes in the leaching environment but rather to physical state of glass surface due to formation of small cracks and new pristine glass areas in contact with water.

Salt cake radioactive waste is a remnant solid salt concentrate after deep evaporation of radioactive evaporator concentrate at WWER NPP’s. The traditional cementing of borate-containing liquid radioactive waste, to which the salt cake belongs, leads to a significant increase in the volume of the final product. This work describes borosilicate vitreous wasteforms developed to immobilize radioactive salt cake waste and comprises data on both glass synthesis and characterization. The composition of glass selected for the purpose of immobilisation of the salt cake radioactive waste allows to include up to 40 wt. % of the oxides contained in the salt cake and to reduce the volume of the final product by more than 2 times compared with the cement compound. The batches were melted in a cold crucible melter at 1200 °C. The normalized cesium leaching rate of the vitrified wasteform product was within range 3.0·10-5 – 3.7·10-6 g/(cm2·day).

It is pertinent to the safety case for geological disposal in the UK that the behaviour of vitrified wastes in proximity to cementitious materials is understood. In this study, vitrified simulant intermediate level nuclear waste (ILW) was subject to dissolution in a synthetic cement water solution to simulate disposal conditions. Results show that the presence of alkali / alkaline earth elements in the cementitious solution can be favourable, at least in the short-term, leading to lower dissolution rates associated with incorporation of these elements into the altered layer of the glass.

Copper is a candidate for use as an overpack material in deep underground nuclear waste disposal. Copper, however, is susceptible to corrosion following closure of the repository and migration of the corrosion products through the buffer material may affect the migration of redox-sensitive radionuclides. Electromigration experiments were performed whereby a copper coupon, which was in contact with compacted bentonite, served as the working electrode and was held at a constant potential of between +100 to +400 mV vs. Ag/AgCl electrode for up to 48 h. The amounts of copper that migrated into the bentonite specimens were found to be in good agreement with the calculated values based on the corrosion current flow for the assumption that copper underwent anodic dissolution as Cu(II). A model based on dispersion and electromigration was able to explain the measured copper profiles in the bentonite specimens. The fitted values of the dispersion coefficient did not depend on the applied potential and were about 10-12 m2/s.

Plasma incineration might be a promising technique for the conditioning of various radioactive waste streams. Assessing the long-term durability of the plasma slag is essential to predict its performance during long-term disposal. In this paper, the stability of six plasma treated surrogate cemented concentrates or resins in a high pH environment is investigated. The slags were crushed (2 different granulometries) and immobilized in a cement matrix, after which samples were submitted to long-term durability tests (stability under water at 20 °C; stability in a high relative humidity environment at 38 °C) and to an accelerated Alkali-Silica-Reaction (ASR) test (1 M NaOH at 80 °C). The first results show that the expansion and strength loss of the cement-slag mixtures remain limited in the test conditions, although differences between the different materials and granulometries could be perceived. No visual damage was observed. Some tests are still ongoing and will last 2 years.

A 1D reactive transport model has been implemented in iCP (interface COMSOL Multiphysics and PhreeqC) to assess the corrosion of Spent Fuel (SF), considered as homogeneous UO2(am,hyd) doped with Pd. The model couples: i) generation of water radiolysis species by alpha and beta radiation considering the complete water radiolysis system with the kinetic reactions involving: H+, OH-, O2, H2O2, H2, HO2-, HO2·, O·, O-, O2-, H·, ·OH and e- ii) processes occurring in the spent fuel surface: oxidative dissolution reactions of UO2(am,hyd) and subsequent reduction of oxidized fuel, considering H2 activation by Pd, and iii) corrosion of Fe(s) in oxic and anoxic conditions. Process i) has been implemented in COMSOL and processes ii) and iii) have been implemented in PHREEQC with their kinetic constants being calibrated with different sets of experimental data published in the open literature. The model yields a UO2(am,hyd) dissolution rates similar to the values selected in safety assessments.

Leaching results to compare the dissolution behavior of a new type of fuel with additives (Advanced Doped Pellet Technology, ADOPT) with standard UO2 fuel are presented. Both fuels were irradiated in the same assembly of a commercial boiling water reactor to a local burnup of ∼58 MWd/kgU. Fuel fragments are leached in simplified groundwater in two autoclaves under hydrogen atmosphere, representing conditions in a canister failure scenario resulting in water intrusion for a spent nuclear fuel repository. Preliminary results indicate the uranium concentration decreased to 3-4x10-8 M after 421 days, slightly above the solubility of amorphous UO2. Xe has been detected in the gas phase of both autoclaves. The concentration of Cs and I seems to gradually approach constant values, yet the redox sensitive elements continue to slowly increase with time. The preliminary data obtained supports the hypothesis that there is no major difference in leaching behavior between the two fuels.

Samples of real irradiated (i-)graphite from bushings of RBMK reactor after several types of decontamination treatments were put in contact with aqueous solutions modelling underground water of Nizhnekansky massif (Russia) equilibrated with bentonite for periods up to 1.5 years. Leach rates of radionuclides and evolution of graphite surface morphology and oxidation state were monitored using nuclear spectroscopy, SEM and XPS. After the experiment, less than one third of surface carbons remains unoxidised. Extensive precipitation of secondary phases (alumosilicates, carbonates) was observed on some samples and is possibly correlated with surface oxidation extent. The leach rates of dose-forming 14C and 36Cl are comparable with the leach rates of radionuclides in glassy waste forms. According to the current study i-graphite could be regarded as a waste form, which is suitable for near-surface disposal.

Alkali-borosilicate glasses (ABS) are used as host immobilization matrices for different radioactive waste streams and are characterized by their ability to incorporate a wide variety of metal oxides with respectively high waste loadings. The vitreous wasteform is also characterized by very good physical and chemical durability. The durability of three ABS compositions were analyzed by investigating their leaching behavior using the MCC1 test protocol and these data were used to investigate the waste components retention in the altered layer and the evolution of the interfacial water composition during the test. The results indicated that the Mg species evolution is exceptional with respect to other alkaline elements and dependent on glass matrix composition and leaching progress, while transition elements speciation is fairly constant throughout leaching process and independent on glass compositions. Si and B species are changing during leaching process and are affected by waste composition. For modified wasteform sample, evolution of Mg, Si and B species is respectively constant, whereas at highest waste loading, these elements have fairly constant speciation evolution within the first 2 weeks of leaching.