Heat generated by radioactive decay of spent fuel represents a potentially important barrier to water accumulation on commercial spent nuclear fuel in breached waste packages. In the absence of water, fuel degradation and radionuclide release will be negligible. Thermal models for the proposed Yucca Mountain Repository suggest that, after a period of approximately 1000-4000 years (depending on loadingand ventilation conditions), the repository drift walls may decline to sub-boiling temperatures, thus allowing humidity in the drift to increase. The question thus arises, is the thermal gradient between the fuel and the drift sufficient to prevent water accumulation in a humid drift environment throughout the regulatory period? The answer depends on the balance between processes that oppose water condensation ontothe fuel (decay heat) and those that promote condensation such as the deliquescence of hygroscopic phaseswithin the fuel.

Our experimental results indicate that deliquescence could lead to the condensation of water onto spent fuel despite the thermal “self-drying”effect if the following criteria are met: (1) the fission product salt CsI is present in the fuel or in the fuel-cladding gap, (2) the relative humidity in the driftexceeds 80% while temperatures in the waste package are around 90oC. Previous work suggests that these criteria may be met for some fuel pins within the proposed Yucca Mountain Repository. However,experiments that account for the role of U(VI) alteration phases suggest that deliquescence may be a self-limiting process in the sense that deliquescent components (e.g. Cs, Ba, Sr) may be incorporatedinto nondeliquescent U(VI) phases that form from the corrosion of spent fuel.

Sand-sized particles of spent nuclear fuel have been found in the environment at Dounreay, north Scotland. These particles, believed to have been discharged during early stages of site operations in the 1960s, have been recovered from areas within the Site, coastal foreshore adjacent to the Site and offshore (marine) environments. As part of the Dounreay Site Restoration Plan, a significant program of work is being undertaken to establish the fate of the Dounreay particles in the marine environment. This program includes materials analysis of particles and particle analogues

The diverse depositional environments, coupled to relatively long fuel-particle residence times, allow insight to be gained into the behaviour of the particles in the natural environment. This includes the effects of physical, chemical and mechanical processes such as particle erosion and corrosion, particle abrasion and particle-saltwater interaction. Using materials analysis techniques including Scanning Electron Microscopy, Energy Dispersive X-Ray Analysis, Electron Probe Microanalysis, together with 137Cs activity measurement we review the fuel-particle mechanical and chemical stability and current radio-isotope composition. These studies allows us to identify the fuel provenance, fuel treatment history and draw some general conclusions of the ultimate fuel particle residence time. We demonstrate that radioisotope modelling strongly indicates that particle break-up is limited by the chemical effects of the marine environment coupled with the native oxide films present on the metallic fuel. Particle transport and distribution are governed by environmental effects on the sediment and sediment transport mechanisms such as tidal currents and storm events.

This study is unique in that it involves the potential effects of long-term saltwater and abrasion interactions with fuel matrixes containing both U-Al and U-Mo of which there is little literature available.

A combination of gamma radiation fields, the absence of moisture, and the high temperatures on the drip shield (DS) and waste package (WP) should combine to suspend microbial activity on the DS/WP surfaces for many tens of thousands of years. This lack of microbial activity, coupled with the corrosion resistance of the titanium Grade7 (Ti-7 drip shield) and the Alloy-22 (waste package) materials make microbially induced corrosion (MIC) of these engineered barrier materials extremely unlikely.

Isolation in a geologic setting has been the generally favored solution to the high-level radioactive waste (HLW) problem since a scientific basis for nuclear waste management began to be formulated over half a century ago. Although general features of suitable settings have been enumerated, quantitative measures of the safety of geologic isolation of HLW are challenging to devise and to implement. Some regulatory measures of isolation for the proposed repository at Yucca Mountain, Nevada, have be devised and revised involving considerations of global releases, groundwater travel time, and time and space scales for isolation. In current Yucca Mountain specific regulations, the measure of long-term safety hinges on probabilistic estimates of radiation doses to the average member of a maximally exposed group of people living about 18 km down the groundwater flow gradient within 10,000 years after permanent closure of the repository. From another perspective, hydrogeochemical studies provide quantitative measures of system openness and the ability of geologic systems to isolate HLW. Hydrogeochemical data that bear on geologic isolation of HLW at Yucca Mountain include precipitation of radionuclides in stable mineralogical products of spent fuel alteration, ages of natural secondary mineralization in the mountain, uranium decay-series isotopic data for system openness, bomb-pulse isotope occurrences, and ambient carbon-14 distributions.

A sample of josephinite, a rock containing predominantly a Ni-Fe metallic phase, was evaluated as a natural metal analog to increase confidence in the assessment of waste package performance for the potential high-level radioactive waste repository at Yucca Mountain, Nevada. The josephinite sample was characterized electrochemically in simulated groundwater environments using cyclic potentiodynamic polarization and potentiostatic tests. The passive surface layers formed potentiostatically were examined by X-ray photoelectron spectroscopy. These results were compared to those obtained with a synthetic Ni3Fe alloy with a chemical composition similar to that of josephinite. Electrochemical studies showed that josephinite exhibited passivity at a slightly higher pH than did the cast Ni3Fe alloy and was found to be slightly more susceptible to pitting corrosion. The passive films formed on the josephinite and the cast Ni3Fe alloy have a duplex structure consisting of an Fe-rich hydroxide outer layer and a Ni-rich oxide inner layer. Results obtained from this study provide an appropriate characterization of the environmental conditions leading to the passivity and localized corrosion of josephinite. Extended persistence of a stable passive film, however, is essential for the long-term stability of the josephinite sample. The relationship between the passive behavior and the formation of alteration layers needs to be established for assessing the survivability of josephinite.

Borosilicate nuclear waste glasses with various amounts of simulated waste elements have been examined by analytical TEM. This preliminary Scanning Transmission Electron Microscopy (STEM) study evaluates the capability of EELS for mapping coordination parameters of the base glass structure, to apply EDX-mapping and HAADF Z-contrast mapping for the characterisation and shape-mapping of common precipitates in nuclear glasses, such as molybdates, and to study irradiation effects.

The results of the studies of uranium valent states in the borosilicate glasses incorporating the components of uranium-containing sludge of Mining and Chemical Combine (MCC, Zheleznogorsks.) is presented in this work. The glasses were made under oxidative and reducing conditions.

The optical spectrophotometry, nuclear gamma-resonance (NGR) and X-ray diffraction (XRD) showed that glasses produced under oxidative conditions are characterized by the presence of only U(6+), while U(4+) in the reducing conditions is present along with U(6+). The ratio U(6+)/to U(4+) varies in depending on the synthesis conditions.

The glass samples synthesized under oxidative conditions were researched at initial solid state. The others synthesized under reducing conditions was dissolved preliminary without distort of uranium valency.

The effect of U(4+)/U(6+) ratio on the uranium leach rates from the glasses has been studied at 90° using MCC-1 test.

Two ceramics for immobilization of a Zr-REE-actinide fraction of high level waste (HLW) based on zirconolite or/and pyrochlore structures with minor brannerite/lucasite, and fluorite-structured dioxide-based solid solution, were synthesized and characterized. The samples were produced by melting of oxide mixtures at 1500 °C followed by controlled cooling for crystallization. Phase compositions of the samples obtained and waste elements partitioning among co-existing phases were investigated in detail using powder X-ray diffraction, scanning electron microscopy with energy dispersive spectroscopy, and transmission electron microscopy. Cerium enters cerianite-based solid solution, lucasite (if present), and to a lesser extent, pyrochlore and zirconolite. Europium and gadolinium enter predominantly zirconolite and pyrochlore. The highest uranium concentrations were found in a uraninite-based cubic solid solution or pyrochlore and zirconolite.

The effect of fabrication processes on the governing mechanical failure criteria for Alloy 22, Ti-Grade 2/7, and Ti-Grade 5/24 has been evaluated through the development of material specific failure assessment diagrams (FAD). The Barsom and Rolfe Charpy V-Notch (CVN) energy to KICempirical relationship was shown to be conservative when compared to experimental KJC data for Alloy 600 and 690, which indicates that the KIC derivation is most likely conservative for Alloy 22 as well. In contrast, the CVN to KICH2 relationship was found to be non-conservative for Ti-Grade 2/7 and Ti-Grade 5/24. The derived KIC value for Alloy 22 was then used to construct a K-based FAD, which indicated that the governing mechanical failure theory was the Tresca criterion for plastic collapse. Both the derived KIC and experimental KQ values were used in the development of the Ti-Grade 2/7 and Ti-Grade 5/24 FADs. The Ti-Grade 2/7 FAD indicated that the governing mechanical failure theory is also plastic collapse while the Ti-Grade 5/24 FADsuggested mixed mode failure processes where neither linear-elastic fracture mechanics (LEFM) nor Tresca can accurately predict.

Confidence in the ability of repository systems to isolate high level wastes from the environment can be strengthened by placing greater reliance on robust designs for the repository system, and by using multiple lines of evidence to demonstrate that parameters, models and predictions developed for the repository are relevant. A particularly useful approach is to demonstrate that models and predictions incorporate processes that have been shown to be important in existing natural systems.

The behaviour and diffusion mechanisms of helium in nuclear ceramics, such as uranium dioxide spent fuel matrix and zirconolite for the specific conditioning of minor actinides, significantly impact the possible evolution of those matrices in interim storage or disposal conditions. In the framework of spent fuel storage studies, the additional diffusion of gas and fission products in uranium dioxidematrix is also an essential aspect of the R&D. Specific experimental studies have been conducted, devoted to thermal diffusion under 1000 C. Data processing methods, lead to helium diffusion coefficient and associated activation energy of 1.05 eV in the zirconolite and 2 eV in UO2. Comparativelywith the uranium dioxide matrix, the helium diffusion coefficient in zirconolite is 1 to 100 million times higher; this parameter will have to be taken into account to dimension the waste form. Diffusion coefficients measurements between 800 C and 1000 C, investigated by SIMS, showed a very slow diffusion of volatile fission products Xe, I, Te and Cs, with coefficients two or three order of magnitude lower than for helium.