Fabricated stainless steel structures are susceptible to stress corrosion cracking (SCC), despite being placed in chloride-containing natural water or humid atmospheres. The present paper describes a model that can define the conditions under which SCC is initiated and propagated, based on analyses of actual SCC incidents induced at welded flanges of cylindrical stainless steel structures.

Whenever the vitrified radioactive waste canister storage conditions deviate from normal and appropriate conditions due to earthquakes or tsunamis, the exposed canisters are expected to suffer SCC within 400 hours to 7 years, according to the analytical results obtained such as degree of sensitization, residual stress distribution, chloride ion concentration, and temperature.

Carbon steel (C-steel) is studied to be the reference material for the metallic components in the high level waste (HLW) repository concepts of several European countries such as France, Switzerland, Belgium.

Electrochemical impedance spectroscopy (EIS) was performed over a period of 7 years, to determine the instantaneous corrosion rate (CR) of carbon steel (C-steel) in contact with clay porewater in diffusive regime. The study was conducted at the Mont Terri underground research laboratory (URL) located in Switzerland. The test chamber was at a depth of 8 m under anoxic conditions at 90°C in a vertical and descending borehole drilled in Opalinus clay (OPA). Microbial and chemical investigations were conducted on porewater in contact with C-steel as well as directly on C-steel surface further to dismantling.

The results showed clearly a decrease of the CR over time followed by a steady state below 1 µm/year. Sulphate and thiosulphate reducing bacteria were observed in porewater and at the metal surface, with a higher concentration of mesophilic and thermophilic bacteria respectively. The metal surface characterizations revealed the presence of magnetite, mackinawite, hydroxychloride and siderite with local traces of oxidizing species such as goethite.

The Supercontainer is the reference concept for the post-conditioning of vitrified high-level radioactive waste and spent fuel in Belgium. It consists of a prefabricated concrete buffer that completely surrounds a carbon steel overpack. In this highly alkaline environment (pH ∼ 13.6) and under normal conditions (i.e. without the ingress of aggressive species), the carbon steel overpack will be protected by a passive oxide film, which is believed to result in very low uniform corrosion rates.

This paper gives an overview of the status of the uniform corrosion, pitting corrosion and stress corrosion cracking behaviour of carbon steel expected during the waste disposal period.

Vitreous materials are the overwhelming world-wide choice for the immobilisation of HLW resulting from nuclear fuel reprocessing due to glass tolerance for the chemical elements found in the waste as well as its inherent stability and durability. Vitrification is a mature technology and has been used for high-level nuclear waste immobilization for more than 50 years. Borosilicate glass is the formulation of choice in most applications although other formulations are also used e.g. phosphate glasses are used to immobilize high level wastes in Russia. The excellent durability of vitrified radioactive waste ensures a high degree of environment protection. Waste vitrification gives high waste volume reduction along with simple and cheap disposal facilities. Although vitrification requires a high initial investment and then operational costs, the overall cost of vitrified radioactive waste is usually lower than alternative options when account is taken of transportation and disposal expenses. Glass has proven to be also a suitable matrix for intermediate and low-level radioactive wastes and is currently used to treat legacy waste in USA, and NPP operational waste in Russia and South Korea. This report is also outlining IAEA activities aiming to support utilisation of vitreous materials for nuclear waste immobilisation.

Iron and nickel oxidation state and coordination in complex sodium aluminophosphate based glasses suggested as potential matrices for immobilization of legacy high level waste currently stored in stainless steel tanks at PA «Mayak» (Ural reg., Russia) were determined by X-ray absorption fine structure spectroscopy (XAFS: XANES/EXAFS). The glasses containing (wt.%) 20-30 Na2O, 6-12 Al2O3, 40-52 P2O5, 2-5 Fe2O3, 1-3 NiO, 0-6 B2O3, 10-15 other waste oxides produced by quenching of their melts were fully amorphous or contained minor Fe and Ni free phases. Fe in the glasses was found to be predominantly trivalent with an average Fe-O distance and a coordination number (CN) in the first shell of 1.94 to 1.97 Å and 5.2 to 5.8, respectively, mostly in octahedral oxygen environment. Ni is divalent in all the glasses and has in the first shell an average Ni-O distance and CN of 1.97 to 2.03 Å and 4.9 to 5.6, respectively. The first shell of both Fe and Ni is somewhat distorted. The second and further coordination shells are weakly appeared exhibiting no clustering and homogeneous distribution of Fe and Ni ions in the glass network. The data on Fe obtained are in good agreement with those from Mössbauer study of same glasses. After annealing glasses were partly devitrified and interpretation of XAFS data is strongly complicated due to Fe and Ni partitioning among crystalline and vitreous phases.

Water degradation of glass waste forms has been studied extensively under a variety of conditions including of bulk glass immersed completely in static or dynamic water. In practice, the vitrified nuclear waste cracks as soon as poured into a container because of differences in thermal expansion coefficients. In addition, in repository the canisters may be only partially immersed in water. Later, water condenses on the surface of glass which corrodes releasing ions. In this work experiments have been performed to understand these effects on the degradation of International Simple Glass (ISG). Simulated cracks were found to develop pitting corrosion in the crack openings when tested by immersing ISG in water. Under load, these pits concentrated stress and grew as large planar cracks inside the glass. The condensation of water on glass surfaces leads to formation of pits and growth of calcium silicate crystals.

The liquidus temperature (T L ) of rare earth (RE) was determined for alumino-borosilicate glasses for treating americium and curium that have been studied previously. Their work covers a wide range of glass composition with various crystalline phases as primary phase. Present work is aimed at understanding the effect of glass composition on T L for waste glasses designed for vitrifying RE oxides wastes. In a sufficiently narrow composition region, this effect can be represented by a first-order model fitted measured T L versus composition data. Test glasses were formulated by varying of component fractions one-at-a-time. The glasses contained SiO2, B2O3, and Al2O3 as glass formers and Nd2O3 with CeO2 as simulated RE waste. Twenty glasses were made to investigate crystallization as a function of temperature and glass composition. The primary crystalline phase was Ce-borosilicate (Ce3BSi2O10), secondary phases were Al-containing crystals (Al2O3 and Al10Si2O19), and crystalline CeO2. A first-order model was fitted to crystal fraction versus glass composition data. Generally, SiO2 and B2O3 tend to suppress crystallization, Al2O3 has little effect, and, as expected, RE components (Nd2O3 and CeO2) promote it. The correlation coefficient, R 2 , was 0.89 for the primary crystalline phase T L as a linear function of composition.

Glasses of the series (mol.%) 40 Na2O, (20-x) Al2O3, x Fe2O3, 40 P2O5 were irradiated with 8 MeV electrons to doses equivalent of 0.1, 0.5, and 1.0 MGy and characterized by FTIR spectroscopy and ESR at room temperature. FTIR spectra of all the glasses consist of strong bands due to O-P-O stretching modes in (PO4)3- and (P2O7)4- units at 1000-1200 cm-1, P-O-P stretching modes at 900-950 cm-1 (νas) and 700-750 cm-1 (νs), and bending modes in the PO4 units. The wavenumber range lower 800 cm-1 has some contribution due to stretching modes in MO4 and MO6 (M = Al, Fe) units. Moreover the bands at 3300-3700 cm-1 and 1550-1650 cm-1 due to stretching and bending modes in both absorbed and structurally bound H2O molecules were present. As irradiation dose increases the bands due to stretching and bending modes in water molecules and M-O-H bonds become stronger and are split. No essential changes with increasing dose were observed within the spectral range of stretching modes of the O-P-O and P-O-P bonds. Irradiation yields phosphorus-oxygen hole centers - PO4 2- (D 5) and PO4 2- (D 6), and PO3 2- ion-radicals (D 2) observable in ESR spectra of low-Fe glasses. At x>5 their responses are overlapped with strong broad line due to Fe(III). On the whole, with the increase in iron content the glass structural evolution decrease.

Sodium carbonate is currently being considered as a wash-out reagent for the removal of the settled solids in the unagitated Highly Active Liquor (HAL) storage tanks at Sellafield. As the settled solids are expected to comprise mainly zirconium molybdate (ZM), this will result in a challenging feed to the Waste Vitrification Plant (WVP) containing high concentrations of both molybdenum and sodium.

In previous studies, it was shown that at high wash-out waste loadings, i.e. 10 – 12 wt% MoO3 incorporation, there was very little tolerance in ‘Ca/Zn’ base glass for extra sodium before the formation of significant separated sodium molybdate salt phase. However, higher amounts of sodium can be accommodated in borosilicate glasses if the wash-out waste loading is reduced. Further studies have now been carried out to investigate the vitrification of more representative calcined waste feeds. Both pure zirconium molybdate (ZM) and blended ZM-reprocessing waste calcines were produced from the appropriate liquor feeds. The maximum waste incorporations of these two calcines in ‘Ca/Zn’ base glass have been determined, along with a complete product quality assessment. This assessment included measuring the bulk density, degree of crystallinity, heat treatment, durability (Soxhlet and PCT), glass transition temperature, and viscosity.

NUMO and JAEA have been conducting a joint research since FY2011, which is aimed to enhance the methodology of repository design and performance assessment in preliminary investigation stage for the deep geological disposal of high-level radioactive waste. As a part of this joint research, we have been developing glass dissolution models which include various processes derived from glass-overpack-bentonite buffer interaction, considering the precipitation of Fe-silicates associated with steel overpack corrosion, and Si transport through altered layer of glass. The objective of this modeling work is to show comprehensively the lifetime of the vitrified waste due to glass matrix dissolution timescales through sensitivity analysis, and to identify the feature/process that most strongly influences the lifetime, and to identify future R&D issues that would help to improve the nuclide transport analysis with confidential value and the safety case in future. The sensitivity analysis suggested that the duration of the glass dissolution might be predicted in the ranges from 3.8×103 to 1.9×105 years. Also, the results indicated that the precipitation of Fe–silicate has the strongest influence on the long-team behavior of vitrified waste.

Samples of 238Pu-doped single-phase ceramics based on cubic zirconia, Zr0.79Gd0.14Pu0.07O1.93, and monazite, La0.9Pu0.1PO4, have been studied by static leach test in distilled water. Before leach test accumulated doses were (in alpha-decays/m3 x 1026): from 1.6 to 1.7 – for cubic zirconia; and 1.0 – for monazite. Despite high radiation damage both phases remained crystalline according to XRD analysis. The results of static leach tests demonstrate the following Pu normalized mass loss (in g/m2, 90°C, 28 days): from 0.3 to 0.7 – for cubic zirconia; and 1.6 – for monazite. These data are discussed in comparison with results of previous leach tests carried out at lower accumulated doses.

EURO-GANEX aims to recycle both major and minor actinides. As the final waste composition is free from actinides, adapted immobilization matrices should be developed. Synroc is a potential wasteform that has proven itself to be efficient in immobilizing high-level wastes (HLW). In this study, a new composition of Synroc, Synroc-Z, is designed and characterized. The key modification is in decreasing the amount of zirconolite phase, which is the main host phase for actinides and increasing the amount of other phases (hollandite and perovskite). As designed the obtained amount of zirconolite is lower than in Synroc-C compositions. Synroc-Z samples were synthesized with a waste loading of 20 wt.% at various temperatures and pressures via hot-pressing to determine the optimum process parameters, which were determined to be 1150-1200°C and 20 MPa, respectively.

Pellets made of pure sodalite blended with commercial glass frit and pellets made of sodalite, glass frit and a mixture of chloride salts, synthesized through dry pressing and subsequent thermal treatment, were evaluated as a potential matrix for confinement of spent chloride salts coming from pyroprocesses. The sodalite pellets were leached at 23°C and 90°C for 28 days, according to the ASTM C1220-10 procedure. Normalized release rates were estimated for the following elements: Li, Na, Al, Si, K, Rb, Cs, Sr, Ba, La, Nd and compared with literature results. SEM investigations, carried out before and after the leaching tests, show dissolution and re-precipitation phenomena at 90°C.

Glass-ceramics were developed initially for the immobilization of miscellaneous Pu-residues at the UK’s Sellafield site from which it was uneconomic to recover Pu for reuse. Renewed interest in the immobilization of a portion of the UK PuO2 stockpile has led to glass-ceramics being evaluated for bulk Pu immobilization. The Nuclear Decommissioning Authority (NDA) in the UK have proposed hot isostatic pressing (HIP) as a potential consolidation technique for the processing of these wasteforms. In this study, zirconolite based glass-ceramics were investigated to determine an optimum formulation. The yield of zirconolite is shown to vary with glass composition and glass fraction, such that a higher Al content favours zirconolite formation. The sample preparation process is discussed to highlight the importance of a high temperature heat-treatment during sample preparation to achieve high quality HIPed wasteforms.

Results of characterization of 238Pu-doped Eu- and La-monazites using single crystal XRD, Raman and XAFS spectroscopy and TEM are presented. It is shown that despite significant accumulated doses (up to 9x1018 α-decays/gram) the Eu-monazite remains a single crystal. Unusual foamy structures are observed by TEM and are interpreted as recrystallisation of domains damaged by recoil U-ions. Partial recrystallisation of the surface material is also supported by Raman and luminescence data.