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

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

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

Electrochemical impedance spectroscopy (EIS) was performed over a period of 7years, to determine the instantaneous corrosion rate (CR) of carbon steel(C-steel) in contact with clay porewater in diffusive regime. The study wasconducted at the Mont Terri underground research laboratory (URL) located inSwitzerland. The test chamber was at a depth of 8 m under anoxic conditions at90°C in a vertical and descending borehole drilled in Opalinus clay(OPA). Microbial and chemical investigations were conducted on porewater incontact with C-steel as well as directly on C-steel surface further todismantling.

The results showed clearly a decrease of the CR over time followed by a steadystate below 1 µm/year. Sulphate and thiosulphate reducing bacteria wereobserved in porewater and at the metal surface, with a higher concentration ofmesophilic and thermophilic bacteria respectively. The metal surfacecharacterizations revealed the presence of magnetite, mackinawite,hydroxychloride and siderite with local traces of oxidizing species such asgoethite.

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

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

Vitreous materials are the overwhelming world-wide choice for the immobilisationof HLW resulting from nuclear fuel reprocessing due to glass tolerance for thechemical elements found in the waste as well as its inherent stability anddurability. Vitrification is a mature technology and has been used forhigh-level nuclear waste immobilization for more than 50 years. Borosilicateglass is the formulation of choice in most applications although otherformulations are also used e.g. phosphate glasses are used to immobilize highlevel wastes in Russia. The excellent durability of vitrified radioactive wasteensures a high degree of environment protection. Waste vitrification gives highwaste volume reduction along with simple and cheap disposal facilities. Althoughvitrification requires a high initial investment and then operational costs, theoverall cost of vitrified radioactive waste is usually lower than alternativeoptions when account is taken of transportation and disposal expenses. Glass hasproven to be also a suitable matrix for intermediate and low-level radioactivewastes and is currently used to treat legacy waste in USA, and NPP operationalwaste in Russia and South Korea. This report is also outlining IAEA activitiesaiming to support utilisation of vitreous materials for nuclear wasteimmobilisation.

Iron and nickel oxidation state and coordination in complex sodiumaluminophosphate based glasses suggested as potential matrices forimmobilization of legacy high level waste currently stored in stainless steeltanks at PA «Mayak» (Ural reg., Russia) were determined byX-ray absorption fine structure spectroscopy (XAFS: XANES/EXAFS). The glassescontaining (wt.%) 20-30 Na2O, 6-12 Al2O3, 40-52P2O5, 2-5 Fe2O3, 1-3 NiO, 0-6B2O3, 10-15 other waste oxides produced by quenchingof their melts were fully amorphous or contained minor Fe and Ni free phases. Fein the glasses was found to be predominantly trivalent with an average Fe-Odistance and a coordination number (CN) in the first shell of 1.94 to 1.97Å and 5.2 to 5.8, respectively, mostly in octahedral oxygenenvironment. Ni is divalent in all the glasses and has in the first shell anaverage 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. Thesecond and further coordination shells are weakly appeared exhibiting noclustering and homogeneous distribution of Fe and Ni ions in the glass network.The data on Fe obtained are in good agreement with those from Mössbauerstudy of same glasses. After annealing glasses were partly devitrified andinterpretation of XAFS data is strongly complicated due to Fe and Nipartitioning among crystalline and vitreous phases.

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

The liquidus temperature (T L) of rare earth (RE) was determined for alumino-borosilicate glasses fortreating americium and curium that have been studied previously. Their workcovers a wide range of glass composition with various crystalline phases asprimary phase. Present work is aimed at understanding the effect of glasscomposition on T L for waste glasses designed for vitrifying RE oxides wastes. In asufficiently narrow composition region, this effect can be represented by afirst-order model fitted measured T L versus composition data. Test glasses were formulated by varying ofcomponent fractions one-at-a-time. The glasses contained SiO2,B2O3, and Al2O3 as glass formersand Nd2O3 with CeO2 as simulated RE waste.Twenty glasses were made to investigate crystallization as a function oftemperature and glass composition. The primary crystalline phase wasCe-borosilicate (Ce3BSi2O10), secondary phaseswere Al-containing crystals (Al2O3 andAl10Si2O19), and crystallineCeO2. A first-order model was fitted to crystal fraction versus glasscomposition data. Generally, SiO2 and B2O3 tendto 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, 40P2O5 were irradiated with 8 MeV electrons to dosesequivalent of 0.1, 0.5, and 1.0 MGy and characterized by FTIR spectroscopy andESR at room temperature. FTIR spectra of all the glasses consist of strong bandsdue to O-P-O stretching modes in (PO4)3- and(P2O7)4- units at 1000-1200cm-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. Thewavenumber range lower 800 cm-1 has some contribution due tostretching modes in MO4 and MO6 (M = Al, Fe)units. Moreover the bands at 3300-3700 cm-1 and 1550-1650cm-1 due to stretching and bending modes in both absorbed andstructurally bound H2O molecules were present. As irradiation doseincreases the bands due to stretching and bending modes in water molecules andM-O-H bonds become stronger and are split. No essential changes with increasingdose were observed within the spectral range of stretching modes of the O-P-Oand P-O-P bonds. Irradiation yields phosphorus-oxygen hole centers -PO4 2- (D 5) andPO4 2- (D 6), andPO3 2- ion-radicals (D 2)observable in ESR spectra of low-Fe glasses. At x>5 their responses areoverlapped with strong broad line due to Fe(III). On the whole, with theincrease in iron content the glass structural evolution decrease.

Sodium carbonate is currently being considered as a wash-out reagent for theremoval of the settled solids in the unagitated Highly Active Liquor (HAL)storage tanks at Sellafield. As the settled solids are expected to comprisemainly zirconium molybdate (ZM), this will result in a challenging feed to theWaste Vitrification Plant (WVP) containing high concentrations of bothmolybdenum 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 littletolerance in ‘Ca/Zn’ base glass for extra sodium beforethe formation of significant separated sodium molybdate salt phase. However,higher amounts of sodium can be accommodated in borosilicate glasses if thewash-out waste loading is reduced. Further studies have now been carried out toinvestigate the vitrification of more representative calcined waste feeds. Bothpure zirconium molybdate (ZM) and blended ZM-reprocessing waste calcines wereproduced from the appropriate liquor feeds. The maximum waste incorporations ofthese two calcines in ‘Ca/Zn’ base glass have beendetermined, along with a complete product quality assessment. This assessmentincluded 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 aimedto enhance the methodology of repository design and performance assessment inpreliminary investigation stage for the deep geological disposal of high-levelradioactive waste. As a part of this joint research, we have been developingglass dissolution models which include various processes derived fromglass-overpack-bentonite buffer interaction, considering the precipitation ofFe-silicates associated with steel overpack corrosion, and Si transport throughaltered layer of glass. The objective of this modeling work is to showcomprehensively the lifetime of the vitrified waste due to glass matrixdissolution timescales through sensitivity analysis, and to identify thefeature/process that most strongly influences the lifetime, and to identifyfuture R&D issues that would help to improve the nuclide transportanalysis with confidential value and the safety case in future. The sensitivityanalysis suggested that the duration of the glass dissolution might be predictedin the ranges from 3.8×103 to 1.9×105years. Also, the results indicated that the precipitation ofFe–silicate has the strongest influence on the long-team behavior ofvitrified waste.

Samples of 238Pu-doped single-phase ceramics based on cubic zirconia,Zr0.79Gd0.14Pu0.07O1.93, andmonazite, La0.9Pu0.1PO4, have been studied bystatic 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 radiationdamage both phases remained crystalline according to XRD analysis. The resultsof static leach tests demonstrate the following Pu normalized mass loss (ing/m2, 90°C, 28 days): from 0.3 to 0.7 – forcubic zirconia; and 1.6 – for monazite. These data are discussed incomparison with results of previous leach tests carried out at lower accumulateddoses.

EURO-GANEX aims to recycle both major and minor actinides. As the final wastecomposition is free from actinides, adapted immobilization matrices should bedeveloped. Synroc is a potential wasteform that has proven itself to beefficient in immobilizing high-level wastes (HLW). In this study, a newcomposition of Synroc, Synroc-Z, is designed and characterized. The keymodification is in decreasing the amount of zirconolite phase, which is the mainhost phase for actinides and increasing the amount of other phases (hollanditeand perovskite). As designed the obtained amount of zirconolite is lower than inSynroc-C compositions. Synroc-Z samples were synthesized with a waste loading of20 wt.% at various temperatures and pressures via hot-pressing to determine theoptimum process parameters, which were determined to be 1150-1200°C and20 MPa, respectively.

Pellets made of pure sodalite blended with commercial glass frit and pellets madeof sodalite, glass frit and a mixture of chloride salts, synthesized through drypressing and subsequent thermal treatment, were evaluated as a potential matrixfor confinement of spent chloride salts coming from pyroprocesses. The sodalitepellets were leached at 23°C and 90°C for 28 days, accordingto the ASTM C1220-10 procedure. Normalized release rates were estimated for thefollowing elements: Li, Na, Al, Si, K, Rb, Cs, Sr, Ba, La, Nd and compared withliterature results. SEM investigations, carried out before and after theleaching tests, show dissolution and re-precipitation phenomena at90°C.

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

Results of characterization of 238Pu-doped Eu- and La-monazites usingsingle crystal XRD, Raman and XAFS spectroscopy and TEM are presented. It isshown that despite significant accumulated doses (up to 9x1018α-decays/gram) the Eu-monazite remains a single crystal. Unusual foamystructures are observed by TEM and are interpreted as recrystallisation ofdomains damaged by recoil U-ions. Partial recrystallisation of the surfacematerial is also supported by Raman and luminescence data.