In its role as the US Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM) Lead Laboratory for Repository Systems, Sandia National Laboratories (SNL) has completed a quantitative assessment of the long-term performance of the proposed Yucca Mountain repository for spent nuclear fuel and high-level radioactive waste. This performance assessment is based on more than two decades of scientific investigations of the engineered and natural systems that comprise the disposal system, and provides estimates of radiation releases from the disposal system and mean annual radiation doses to a hypothetical “reasonably maximally exposed individual” for one million years following closure of the facility, considering releases from all pathways. This presentation will summarize the technical basis for the performance assessment and review the results presented to the US Nuclear Regulatory Commission (NRC) as part of the demonstration of compliance with applicable long-term radiation standards. Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes. The statements expressed in this article are those of the authors and do not necessarily reflect the views or policies of the United States Department of Energy or Sandia National Laboratories.

The UK has recently established its first National Nuclear Laboratory (NNL). This has been formed out of the Nexia Solutions organisation, formally the Research and Technology subsidiary of BNFL. Over the next year the NNL will develop so that it can fulfil its mission, which will include the development and maintenance of key skills and undertaking strategic R&D programmes both in the UK and in international collaborations. A key role of the NNL will be to enhance its interactions with Universities to facilitate skills transfer into the nuclear industry as well as support its R&D programmes. Over the past decade the NNL and its predecessor has already established close relationships with leading Universities in the UK including Sheffield, Manchester, Leeds and Imperial College London. This paper will describe the future plans for working with Universities as the NNL develops, building on the success to date.One of the key objectives of the NNL has been to work in an integrated way with University researchers with programmes spanning fundamental research through to applied R&D.The Immobilisation Science Laboratory (ISL) at Sheffield University was established by the NNL predecessors and a range of joint research has been undertaken over the past 8 years. This includes support for the UK Pu disposition programme where R&D has included investigating a range of ceramic and glass wasteforms which has allowed a number of ceramic wasteforms to be down selected for detailed evaluation. Cementation research has included understanding wasteform performance, for example long-term durability. This has involved work on determining free water and the implications for immobilisation of reactive wastes.Other work involving the NNL and Universities, in particular at Leeds and Manchester, has considered the characteristics and behaviour of intermediate and high level waste sludges. This has included determining the chemical speciation of actinides and fission products, and physical properties of active and simulated sludges using experimental and modelling techniques.A significant programme on environmental work has been undertaken by the NNL. In research applicable to low level waste disposal and contaminated land reactive transport modelling is utilised to apply experimental and field based research in environmental geochemistry and radiochemistry undertaken by university collaborators. This includes research into Sr-90 interactions with wastes and soils and synchtrotron studies of green-rust corrosion products and their potential to uptake contaminants. This paper will describe a number of examples of R&D carried out by NNL in association with its University partners.

Direct geological disposal has for a long time being considered in many countries as a reference solution to ultimately manage Spent Nuclear Fuel (SNF). However, the recent concerns about the global climate change and the strong increase of energy demands in the world leads to a remarkable renaissance of nuclear energy for a few years with anticipated new reactor and plant construction. In this new context, resources have to be preserved and recycled as far as possible. Directly disposing SNF in deep underground may not be the most sustainable solution since 96% is still recyclable (U, Pu, minor actinides). Recycling part of the actinides is therefore an option of growing interest for many countries instead of direct disposal. This significant policy evolution has obviously to influence the relative importance of the different scientific research fields, in particular regarding SNF. This papers aims to depict the scientific key issues related to the different options considered for managing spent nuclear fuel, direct disposal and recycling.Studies on SNF long term evolution in direct disposal has significantly developed in the last decade. They allow deriving reliable and scientifically-sounded radionuclides source term to be used in safety analyses, especially for the Instant Release Fraction and the radiolytic dissolution which is more deeply understood. However, key scientific areas are still to be addressed, in particular on the IRF inventory and secondary phases precipitation.Regarding the closed cycle scenario, spent nuclear fuel is supposed to be ultimately digested in highly acidic conditions. SNF dissolution properties have to be significantly improved. However, SNF is also probably to be stored for some time in order to optimise the global U and Pu stockpiles in the whole cycle (to avoid any Pu accumulation). In long term storage, questions regarding long term cladding performance are of importance for the safety demonstrations. In particular, long term creep and rupture criteria in dry storages, as well as cladding embrittlement and potential rupture in wet storage (French choice) have to be accurately defined. Partitioning actinides is also a field of significant research area in order to improve the robustness and proliferation-resistance of the current processes and allow future recovering of minor actinides (MA). After partitioning of U, Pu +/- MA, long-lived radionuclides are confined by a dedicated matrix, the nuclear glass. Its performance has been proven to last long enough (> 300ky.) to allow a safe disposal in reducing deep geological environment. Main improvements are currently anticipated in the field of (i) the confinement capabilities of nuclear glass (higher alpha loading …) and (ii) elucidation of glass dissolution / environment interactions in the geological disposal. In conclusion, this paper will focus the main scientific inputs which are needed to support the global optimisation of the SNF sustainable cycle.

Understanding the corrosion of spent nuclear fuel (SNF) and the subsequent mobilization of released radionuclides, particularly under oxidizing conditions, is one of the key issues in evaluating the long-term performance of a nuclear waste repository. However, the large amounts of iron in the metal waste package may create locally reducing conditions that would lower corrosion rates for the SNF, as well as reduce the solubility of some key radionuclides, e.g., Tc, Np, and U. In order to investigate the interactions among SNF-waste package-fluids, six small-scale (~1:40 by length) waste package mockups were constructed using metals similar to those proposed for use in waste packages at the proposed repository at Yucca Mountain (YMR). Each mockup experiment differed with respect to water input, exposure to the atmosphere, and temperature, and two of the mockups contained 0.1 g of UO₂. Simulated Yucca Mountain process water (YMPW) was injected into five of the mockups at a rate of 200 μL per day for five days a week using a calibrated needle syringe. The YMPW was prepared by equilibrating 50 mg/L silica as sodium metasilicate with air and adding enough HCl to lower the pH to 7.6 in contact with an excess of powdered calcite. X-ray powder diffraction, scanning electron microscopy, and electron microprobe analysis confirm that, despite interactions with air outside the waste package, the dominant corrosion product in all cases was the Fe(II)-bearing magnetite. In the high temperature (60 degrees celsius) experiment, hematite and a fibrous, Fe-O-Cl phase were also identified. The Fe(II)/Fe(III) ratios measured in the corrosion products using a wet chemistry technique indicate extremely low oxygen fugacities (10^-36 to 10^-59 bar). The uranium observed associated with the corrosion products is still UO₂, suggesting that even after two years conditions were sufficiently reducing to minimize oxidation. The small amounts of U that were dissolved and removed from the original grains can be found in the water (0.5-5 ppb U) and associated with the steel corrosion products surrounding the UO₂ grains. Although these experiments were two years in duration, they still do not address the long-term behavior of a breached waste package; however, these results do support the possibility of a transient period of reducing conditions within a breached waste package.

In this experimental work, the corrosion behaviours of SF and canister materials (cast iron and copper), as well as immobilization reactions of radionuclides on iron canister material were investigated under simulated early canister failure conditions.The leaching solution, water with 10 mM NaCl and 2 mM NaHCO3, was deoxygenated by bubbling with pre-deoxygenated 99.97%Ar + 0.03% CO2 gas mixture before and during the experimental period. The gamma-radiation level experienced by the leaching system in the hot cell during the whole experimental period was 850 mGy / hour. During the initial leaching period, the concentration for all radionuclides dissolved from a SF pin increased with time, but after introducing small foils of Fe-Cu-cast iron in the leaching solution the concentration of U, Tc and Np dropped sharply, suggesting the precipitation of insoluble oxides on the surface of iron foils. pH, Eh and Ecorrosion for Fe, Cu and cast iron were recorded during the whole experiment. The polarization resistance Rp was measured. The corrosion rates of iron, cast iron and copper were calculated based electrochemical measurement and also estimated from the thickness of the corrosion layers after two years of corrosion. Similarities in Ecorr values and corrosion product layers (non-uniform, 20 - 50 micro m thick after two years) for pure Fe and cast-iron, indicate that the two materials exhibited similar corrosion behaviour. The corrosion rate of copper was found to be 200 times slower than that measured under air saturated conditions.An iron-silica rich corrosion layer containing 1-2 micro m sized U-Si rich particles was observed by SEM-WDS analysis on the cross section of reacted iron foils. Trace nuclides, Np, Pu, Tc, Sr and Cs were also detected by sensitive SIMS method.The effects of adding 10% H2 in the purging gas mixture on the SF leaching and radionuclides sorption on glass vessel were also investigated. The ratios of various radionuclides in the leaching solutions at end of the experiments and the corresponding radionuclides sorbed/precipitated in glass vessel wall were interpreted as consequences of redox and sorption/desorption reactions.After the leaching experiments, the SF pin was reacted in a saturated Fe(II) solution (FeCO3) solution for 450 days. The formation of ferric precipitates on SF surface was observed. The leaching rates of all radionuclides were found to be largely decreased as compared to the initial leaching experiment.Information obtained in this work is useful for discussing the behaviour of SF and canister materials under early-canister-failure conditions at deep geological repositories.

Spent nuclear fuel and other high level nuclear wasteforms are subjected to radiation damage. Fission damage accumulated during in-pile irradiation amounts to thousand of displacements per atom (dpa). Spent fuel also accumulates alpha-decay damage and He during storage. The dose rates and the temperatures experienced in this case are lower than for in-pile operation, and depend on composition, history and activity of the fuel: however, the duration of the storage is longer (up to a few hundred years if extended interim storage concepts are considered); if final disposal in the repository is considered, the time interval in which radiation damage accumulates is open-ended.This presentation shows highlights from studies on irradiated fuels and analogue materials (both natural and tailor-made). The effects due to accumulating alpha decay damage and He on the corrosion resistance and overall durability of spent fuel will be discussed. The final goal of these studies is to assess the long-term stability of spent fuel.In order to simulate long-term accumulation of alpha-decay damage within timeframes suitable for laboratory experiments, alpha-doped materials can be used, i.e. matrices loaded with short-lived alpha-emitters (like e.g. Pu-238, U-233). The evolution of microstructural defects and corresponding macroscopic properties as a function of accumulated dose for spent fuel and alpha-doped UO₂ shows remarkable similarity of behaviour. The activity levels for possible dose rate effects (or artefacts) to occur will be discussed. Experiments combining annealing studies using calorimetric techniques and He-release as a function of temperature with microstructure examination using TEM were performed to investigate the correlation among the annealing of defects in the microstructure, the release behaviour of He, and the heat effects associated with these processes in the material.Alpha-doped UO₂ containing different fractions of alpha-emitter simulates the level of activity of spent fuel after different storage times, and can be used to study the effects of water radiolysis on the corrosion behaviour of aged spent fuel exposed to groundwater in a geologic repository. Alpha-doped UO₂ with specific activities spanning over five orders of magnitude was used in static leaching experiments at room temperature in deionized water or in groundwater under various nominal redox conditions and using different Surface/Volume (S/V) ratio. Within the context of the experimental conditions considered, these experiments allowed establishing significance and extent of the alpha-radiolysis enhancement of fuel dissolution. The radiolytic enhancement of the corrosion process was measured for the alpha-doped materials compared to undoped UO₂. At relatively low S/V conditions, an activity dependence of the concentration levels in the leachate was evident, while saturation and precipitation of oxidized phases characterized the evolution of the leaching tests at high S/V.

An accident scenario for nuclear spent fuel dry storage consists in cask and fuel rod simultaneous failures that will put nuclear ceramic in contact with air. The swelling associated to UO2 oxidation in U3O8 might lead to the rod ruin. In literature interpretation, U3O8 is associated to the sigmoid part of the UO2 oxidation weight gain curve, and its 36% crystalline swelling is taken for responsible of the rod degradation. That is why previous criterions for safe behaviour of defected fuel rod in contact with air took into account U3O8 molar fraction. Recently the sigmoid part of the UO2 oxidation weight gain curves was reinterpreted as partially resulting from ceramic cracking and increased reactive surface; ceramic cracking induces sample bulking leading up to 300% swelling, one order of magnitude higher than the crystalline swelling induced by the UO2 to U3O8 transformation. This new interpretation led us to propose a new criterion of safety in which no significant damage is expected in the defective spent fuel rod before cracking apparition. The apparition of significant cracking and consecutive bulking is associated to the formation of a critical depth of U4O9+γ the time, at which the U4O9+gamma critical depth is formed, is calculated thanks to a finite difference model derived from a previously published model dedicated to un-irradiated UO2 oxidation. Although this new criterion significantly reduces the margin before cladding damage because cracking occurs before U3O8 formation, it leads to similar evaluations for the duration of safety than some previous ones.

The release of toxic and radioactive species from spent fuel in contact with water is expected to depend mainly on the rate of dissolution of the UO₂ matrix. In Sweden, the spent fuel will be disposed at 500-700 m below ground level, where the conditions are reducing and UO₂ has very low solubility. However, ionizing radiation emitted from the spent fuel will produce oxidants (and reductants) and alter the otherwise reducing conditions and thus enhance the rate of spent fuel dissolution. The burn-up of future spent fuel to be disposed will be higher than the burn-up of today’s fuel. Actinides accumulate in the rim zone and the content of lanthanides and other fission products will also increase in spent fuel as a consequence of higher burn-up. Furthermore, the formation of metallic particles will be enhanced. The increased actinide content in spent fuel at higher burn-ups will lead to a higher α-dose rate in the surrounding water and the higher content of fission products will also contribute to a higher β- and γ-dose rate initially. The dissolution rate is expected to increase with higher burn-up due to higher dose rates. On the other hand, the presence of fission products like lanthanides in the UO₂ matrix has been shown to have an inhibiting effect on UO₂ dissolution. The outcome of the study will be discussed taking the above mentioned as well as other physical properties into account. Previous static corrosion tests on spent fuel with a burn-up range of 27 to 49 MWd/kg U showed that the cumulative release fractions increase with burn-up to reach a maximum at approximately 40-45 MWd/kg U. At higher burn-up (up to 49 MWd/kg U) the release rates decrease. The study has now been extended to comprise spent fuel with even higher burn-up. Static leaching of spent fuel with 60 and 75 MWd/kg U burn-up has been started. From each spent fuel rod, a fuel pin segment, containing one complete and two half pellets, is leached under oxidizing conditions in synthetic groundwater. Results from five contact periods, for a cumulative contact time of one year, will be presented and compared with previous results.

Understanding the mechanism of aqueous corrosion of nuclear waste borosilicate glasses is essential to reliably predict their long-term aqueous durability in a geologic repository. Here we report the results of corrosion experiments with borosilicate glass cuboids with edge lengths of about 2.5 mm in a HCl solution of pH = 0 at 150°C for 6 to 336 hours, including two experiments with ¹⁸O and ²⁶Mg as isotope tracers. Several observations were made in this study such as (i) the occurrence of chemical oscillations in the corrosion rim that is composed of silica, (ii) an enrichment of ¹⁸O and ²⁶Mg in the corrosion rim formed in the isotopically enriched solutions without observable diffusion profiles, (iii) a sharp phase boundary of the corrosion rim towards the pristine glass, (iv) a high porosity in the corrosion rim, and (v) the occurrence of silica spherules at the surface while the overall shape of the cuboids was retained. These features are not at all compatible with classical theories about the formati