The Yucca Mountain repository combines multiple barriers, both natural and engineered, which work both individually and collectively to limit the movement of water and the potential release and movement of radionuclides to the accessible environment. Engineered structures, systems and components (SSCs) are designed to function in the natural environment utilizing materials chosen to perform their intended functions in order to meet the postclosure performance objectives. Similarly, the features of the natural environment are expected to respond to the presence of the repository through geomechanical, hydrogeologic and geochemical changes. At Yucca Mountain, specific features, both engineered and natural have been identified as requiring design control during repository construction and operations. The integration between design and postclosure safety analysis is facilitated using design control parameters. The term “design control parameters” includes functions and performance requirements allocated to SSCs through the design process, as well as the attributes of SSCs that are developed during design (e.g., dimensions; weights; materials; fabrication and quality-control processes; and operating conditions). These control parameters provide an interface between the design and the analyzed postclosure safety bases, which needs to be maintained through the licensing process. Maintenance of the design is controlled through configuration management and procedural safety controls. The design control parameters serve three key purposes. First, they identify key aspects of the design that serve as the design bases for the designers of the SSCs of the facility. Second, they provide a useful input to the analyses of relevant postclosure features, events and processes (FEPs) and are used to either exclude FEPs from the postclosure safety analysis or as an input to models of included FEPs in the safety analysis. Finally, they provide important controlled interface constraints between the design and safety analyses organizations that are amenable to configuration management. Several examples of such design controls will be presented in this briefing. The first type of design controls relates to the location of the underground facility, including standoffs from faults and the ground surface. The second type of design controls relates to the configuration of the engineered features including the spacing of emplacement driftsand drip shield dimensions and characteristics. A third type of design controls relates to constraints on handling, loading and emplacing waste forms in canisters and waste packages in the emplacement drifts.

Concentrations of Ra in groundwater was thought to be affected by substitution reactions with alkaline earth elements contained in a mineral, because the concentration of Ra is less than the solubility of known Ra containing phases. The substitution reaction was simulated using a partition coefficient. Calcite is a dominant mineral to react with Ra in a geological disposal system. However, previous identification of reactive layers of calcite or reversibility of the substitution reaction has not been confirmed. A re-distribution experiment was therefore undertaken and it was found that an estimated 21±13 layers near the surface were reactive and within them substitution equilibrium was achieved. Using these results, a model to estimate Ra concentration was established and adopted to analyze Ra migration. The effects of substitution of Ra and Ca were reasonably simulated.

The influence of boron on the kinetics of oxidation of iron in silicate melts relevant to nuclear waste storage has been investigated by XANES experiments. The measurements have been performed isothermally as a function of time at the iron K-edge. The redox kinetics become slower with increasing B2O3 content either close to the glass transition range, where the redox kinetics are controlled by diffusion of network-modifying cations, or at superliquidus temperatures where oxygen diffusion is the rate-limiting factor. In both ranges the kinetics can be interpreted in terms of boron speciation and interaction with alkali cations. Below the liquidus, however, the long times needed to reach redox equilibrium allow sintering of the powders investigated to take place so that the resulting changes in sample geometry prevent determinations of oxidation kinetic parameters from being made.

Experimental results using environmental SEM on intentionally defected fuel particles showed that oxidation induced cracking could lead to the degradation of HTR coated particles. The interpretation proposed for the swelling resulting from cracking can be extended to irradiated nuclear fuels. That is why a new criterion was proposed to defined safe handling of defective fuel in dry storage condition. This criterion defines the time needed to create an oxidized layer thickness leading to significant cracking.

The presence of significant quantities of fluoride and chloride in four types of legacy wastes from plutonium pyrochemical reprocessing required the development of a new wasteform which could adequately immobilize the halides in addition to the Pu and Am. Using a simulant chloride-based waste (Type I waste) and Sm as the surrogate for the Pu3+ and Am3+ present in the waste, AWE developed a process which utilised Ca3(PO4)2 as the host material. The waste was successfully incorporated into two crystalline phases, chlorapatite, [Ca5(PO4)3Cl], and spodiosite, [Ca2(PO4)Cl]. Radioactive studies performed at PNNL with 239Pu and 241Am confirmed the process. A slightly modified version of the process in which CaHPO4 was used as the host was successful in immobilizing a more complex multi-cation oxide–based waste (Type II) which contained significant concentrations of Cl and F in addition to 239Pu and 241Am. This waste resulted in the formation of cation-doped whitlockite, Ca3-xMgx(PO4)2, β-calcium phosphate, β-Ca2P2O7 and chlor-fluorapatite rather than the chlorapatite and spodiosite formed with Type I waste.

Development work has been carried out for several years on the cementation of phosphate waste from the Tokai Reprocessing Plant Solvent Waste Treatment Facility. The phosphate waste consists of mainly NaH2PO4 at a concentration of approximately 440g/ℓ, and has a pH of approximately 4. During attempts to encapsulate this waste in cement the phosphate species interfere with the hydration reaction and retard the setting of the cement and reduce the strength of the wasteform. To mitigate these detrimental effects, pretreatment of the phosphate waste with Ca(OH)2 prior to cementation has been investigated at small scale. The effect of pre-treating the waste with various amounts of Ca(OH)2, specified by the Ca(OH)2/ NaH2PO4 molar ratio, was investigated under a range of mixing conditions. The viscosity of the pre-treated waste simulant was recorded and an optical microscope was used to identify the presence of small crystals formed during pre-treated. Cementation of pre-treated phosphate waste simulant was also investigated using various conditions to assess the effects of changing the waste loading, mixing temperature and water/cement ratio. The dimensional stability and strength of the resulting cemented waste was assessed up to 28 days. Results from the trials show that pre-treatment of the waste simulant with Ca(OH)2 is particularly sensitive to the Ca(OH)2/ NaH2PO4 molar ratio and mixing temperature. These two factors strongly affect the rheology of the pre-treated waste and the acceptability for subsequent cementation. An optimum Ca(OH)2/ NaH2PO4 molar ratio was found to exist and the importance of controlling the waste temperature during pre-treatment was explained. Cementation trials were performed using the pre-treated phosphate waste simulant and slag cement. It has been shown that pre-treated phosphate waste may be encapsulated in slag cement where the wasteform achieved compressive strengths exceeding 10 MPa after 28 days curing at waste loadings exceeding 13wt% (given as a % weight of cement wasteform represented by the NaH2PO4 in the untreated waste). Demonstration trials of pre-treatment and encapsulation will be executed at full scale in the future.

Advanced nuclear fuel reprocessing can partition wastes into groups of common chemistry. This enables new waste management strategies not possible with the plutonium, uranium extraction (PUREX) process alone. Combining all of the metallic fission products in an alloy and the balance as oxides in glass minimizes high level waste (HLW) volume. Implementing a waste management strategy using state-of-the-art combined waste forms and storage to allow radioactive decay and heat dissipation prior to placement in a repository makes it possible to place almost 10x the HLW equivalent of spent nuclear fuel (SNF) in the same repository space. However, using generic costs based on preliminary studies for waste stabilization facilities and separations modules, this analysis shows that combining the non-actinide wastes and using only one glass waste form is the most cost-effective.

The effect of salt concentration on swelling pressure of water-saturated bentonite was modelled based on the thermodynamic data of water at montmorillonite surface and of water in solutions of various salinities coming in contact with the montmorillonite. Activities of water at montmorillonite surface were obtained as a function of water content and temperature at montmorillonite densities of 0.6 and 0.9Mg/m3 by a vapor pressure method, and the relative partial molar Gibbs free energies (dGw) were determined. Water affected from the montmorillonite surface was estimated to be almost all interlayer water. Swelling pressure vs. montmorillonite partial density was estimated for solutions of various salinities and compared to data measured for various montmorillonite contents and different silica sand contents of bentonites. The calculated swelling pressure clearly decreased with increasing salinity. The effect of salinity on swelling pressure was not clear in the measured data and was within the range of the scattering in the measured data. This cause is presumed to be due to the increase of the ionic strength of porewater by dissolution of soluble minerals in the bentonite.

Non-cementitious grouts have been tested in Olkiluoto for the sealing of fractures with the small hydraulic aperture. A promising non-cementitious inorganic grout material for sealing the fractures of the apertures less than 0.05 mm is commercial colloidal silica called silica sol. The use of colloidal material has to be considered in the long-term safety assessment of a spent nuclear fuel repository. Objective of this work was to determine colloid release from the silica sol gel and stability of silica colloids in different groundwater conditions. To use silica sol as a grout, the injected colloids have to aggregate and form a gel within a predictable time by using a saline solution as an accelerator. Silica sol gel samples were stored in contact with medium salinity and low salinity groundwater simulates. Release of silica colloids and colloid stability was followed by analyzing the colloid concentration, particle size distribution, concentration of reactive silicon, solution pH and zeta potential after one month, half a year and one year. Malvern Zetasizer Nano ZS equipment was used to determine colloidal particle size distributions applying the dynamic light scattering method and zeta potential based on dynamic electrophoretic mobility. The colloidal particle concentration was estimated from Zetasizer measurements applying a standard series. Dissolved reactive silica concentration was determined using the molybdate blue method and total silica concentrations were determined using ICP–MS. The release and stability of silica colloids were found to be dependent significantly on groundwater salinity. Zeta potential values near zero and the increase in particle size at first and then the disappearance of large particles indicated particle flocculation or coagulation and instable colloidal dispersion in a saline groundwater simulate. In low salinity ground water simulate high negative zeta potential values, small particle size and constant size distribution indicate the existence of stable silica colloids. The concentrations of the released colloids were slightly higher than determined in natural granitic ground waters. Under prevailing saline groundwater conditions in Olkiluoto no significant release of colloids from silica sol is expected but the possible influence of low salinity glacial melt waters has to be considered.

Alloy 22 (N06022) is a highly corrosion resistant nickel based alloy. Extensive research has been conducted in the last eight years on the corrosion behavior of Alloy 22, mainly regarding its resistance to localized corrosion. Less attention has been paid to the general corrosion resistance in highly concentrated brines that may result from the deliquescence of salts contained in dust. Salts such as mixtures of NaCl, KCl, CaCl2, NaNO3, and KNO3 may deliquesce at temperatures above 100°C through absorption of moisture from the air. Electrochemical tests were used to assess the general corrosion behavior of Alloy 22 in brines with chloride and nitrate concentrations ranging from 8 molal to 100 molal in the temperature range 100 to 160°C. The effect of mixed anions and cations was also studied. Results show that, even for short-term immersion periods, the corrosion rate of Alloy 22 in high temperatures super concentrated brines is generally below 10 μm/year.

The Immobilisation Science Laboratory, University of Sheffield, is working with Magnox South Ltd to develop a range of glass formulations that are suitable for vitrification of the Wet Intermediate Level Waste (WILW) envelope arising from decommissioning of the Hinkley Point ‘A’ (HPA) power station. Four waste mixtures or permutations are under consideration for volume reduction and immobilisation by vitrification. The inorganic fractions of several of the wastes are suitable for vitrification as they largely consist of SiO2, MgO, Fe2O3, Na2O, Al2O3 and CaO. However, difficulties may arise from the high organic and sulphurous contents of certain waste streams, particularly spent ion exchange (IEX) resins. IEX resin wastes may be the key factor in limiting waste loading, and possible thermal pretreatments of IEX resin to decrease C and S contents prior to vitrification have been investigated. Our results suggest that lowtemperature (90 °C) pretreatment is more favourable than hightemperature (250, 450, 1000 °C) pretreatment. A thorough desktop study has provided initial candidate glass compositions which have been downselected on the basis of glass forming ability, melting temperature, viscosity, liquidus temperature, chemical durability and potential sulphate capacity. Early results for two of the candidate glass formulations indicate that formation of an amorphous product with at least 35 wt % (dry waste) loading is achievable for HPA IEX resin wastes.

After closure, the near-field of a nuclear waste repository contains large amounts of oxygen in tunnels and deposition holes. The bentonite buffer/backfill will contain oxygen as a gas phase in unsaturated pores as well as dissolved gas in porewater. The redox conditions in the bentonite filling after post-closure will change towards reducing conditions. In the initial stage, the development of the redox state is mainly governed by the depletion of oxygen. The main mechanisms of oxygen depletion in the bentonite are: 1) diffusion into the surrounding rock and 2) reactions with accessory minerals and by microbial aerobic consumption of organic matter [1,2]. The reactions leading to oxygen depletion are not, however, well understood. The objective of this work was to gather new information concerning oxygen depletion in MX-80. This was done by measuring oxygen depletion and changes in the redox state in suspensions of 1) MX-80, 2) a heavy fraction of MX-80, or 3) a light fraction of MX-80.

Soft X-ray spectroscopy was applied to study a calcium bentonite from the Kutch area in India. We recorded the X-ray absorption spectra from the L-edge of calcium, silicon, and aluminum, and from K-edge of oxygen. The Ca absorption spectrum shows a quasi-atomic behavior, while the Si spectrum closely simulates the absorption spectrum of a pure silicon oxide. The O K spectrum shows a pre-peak, which is absent in the spectra of both the pure, bulk aluminum and silicon oxides. The Al L spectrum is complex and shows almost no resemblance to the absorption spectrum of aluminum oxides. The chemical state of the Al atoms (in octahedral coordination) must, thus, be quite different from what is common in the oxides. The obtained data show that soft X-ray spectroscopy is a promising technique for studying clay minerals. It is capable of supplying unique information that is complementary to information accessible using other techniques; especially, it can be used to determine the local electronic structure at various atomic sites in the complex samples.

This work demonstrates that absorption near-edge structures of the O K-edge in zircon are directly related to atom structure, including bond length, bond angle and coordination. Small variations in the structure can induce measurable changes in absorption spectrum.

Evaluation of the transport and retardation properties of rock matrices that serve as host rock for nuclear waste repositories necessitates their thorough pore-space characterization. Relevant properties to be quantified include the diffusion depth and volume adjacent to water conducting features. The bulk values of these quantities are not sufficient due to the heterogeneity of mineral structure on the scale of the expected transport/interaction distances. In this work the 3D pore structure of altered granite samples with porosities of 5 to 15%, taken next to water conducting fractures at 180 200 m depth in Sievi, Finland, was studied. Characterization of diffusion pathways and porosity were based on quantitative autoradiography of rock sections impregnated with C14-labelled polymethylmethacrylate (PMMA). Construction of 3D structure from PMMA autoradiographs was tested. The PMMA method was augmented by field emission scanning electron microscopy and energy-dispersive X-ray analyses (FESEM/EDAX) in order to study small pore-aperture regions in more detail and to identify the corresponding minerals. The 3D distribution of minerals and their abundances were determined by X-ray microtomography. Combining the mineral specific porosity found by the PMMA method with these distributions provided us with a 3D porosity distribution in the rock matrix.

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. 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. This paper describes the future plans for working with universities as the NNL develops, building on the success to date. Joint R&D research programmes already cover many important areas for the nuclear industry. This includes support for the UK Pu disposition programme where R&D at the Immobilisation Science Laboratory, Sheffield 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 paper will describe a number of examples of R&D carried out by NNL in association with its university partners.

Magnesium oxide (MgO) is the only engineered barrier certified by the EPA for emplacement in the Waste Isolation Pilot Plant (WIPP), a U.S. Department of Energy repository for transuranic waste. MgO will reduce actinide solubilities by sequestering CO2 generated by the biodegradation of cellulosic, plastic, and rubber materials. Demonstration of the effectiveness of MgO is essential to meet the U.S Environmental Protection Agency's requirement for multiple natural and engineered barriers. In the past, a series of experiments was conducted at Sandia National Laboratories to verify the efficacy of Premier Chemicals LLC (Premier) MgO as a chemical-control agent in the WIPP. Since December 2004, Premier MgO is no longer available for emplacement in the WIPP. Martin Marietta Magnesia Specialties LLC is the new MgO supplier. MgO characterization, including chemical, mineralogic, and reactivity analysis, has been performed to address uncertainties concerning the amount of reactive constituents in Martin Marietta MgO. Characterization results of Premier MgO will be reported for comparison. Particle size, solid-to-liquid ratio, and stir speed could affect the rate of carbonation of MgO slurries. Thus, it's reasonable to hypothesize that these factors will also affect the rate of hydration. Accelerated MgO hydration experiments were carried out at two or three levels for each of the above factors in deionized water at 70 °C. The Minitab statistical software package was used to design a fractional-factorial experimental matrix and analyze the test results. We also fitted the accelerated inundated hydration data to four different kinetic models and calculated the hydration rates. As a result of this study we have determined that different mechanisms may be important for different particle sizes, surface control for large particles and diffusion for small particles.

This paper focuses on how to extrapolate current knowledge of spent fuel matrix alteration processes from laboratory to repository conditions, i.e., the influence of changes in both the initial surface oxidation level and the evolution of the specific surface area during the alteration process. Therefore, a spent fuel matrix alteration model allowing the alteration rate evolution to be predicted as a function of both the host rock considered and evaluation time scale of interest is described. At present, the model assumes that alteration of the spent fuel will start when the groundwater reaches the solid surface and that only the radiolytic species of the groundwater (oxidants generated by a-radiation of spent fuel) will produce the surface oxidation process and subsequent matrix dissolution; O2, H2O2 and OH are the species that react with UO2(s) for oxidation of the pellet surface. The dissolution process of the surface sites that are oxidized is modelled in two steps: first, a surface co-ordination of the oxidized layer with aqueous ligands and, second, detachment (dissolution) of the product species. Taking this mechanism into account, the model gives the evolution of the spent fuel matrix alteration rate over periods as long as 1,000,000 years. In this work is focussed on input the experimental results obtained of UO2 surface area behaviour (presented in previous MRS conference), on the MAM model. The matrix alteration rate results obtained, with MAM model, for repository granitic environment will be presented and compared to those performed for SFS project. Furthermore, a sensitivity analysis study has been performed on the influence of the following variables: Influence of the initial power size distribution and the initial oxidation state

This study addresses the diffusion of representative sorbing elements, cobalt, cesium and europium in the Opalinus Clay (OPA). The methodology used here to determine diffusion coefficients is the ‘instantaneous planar source’ method. In this setup, a paper filter impregnated with tracer is introduced between two clay samples, avoiding contact between the tracer and the experimental vessels. The apparent diffusion coefficients (Da) perpendicular to the bedding plane, obtained with this experimental method and fitting the experimental results with an analytical solution, were Da(Co) = (2.4-3.5)·10-14 m2/s, Da(Cs) = (5.9-8.0)·10-14 m2/s, and Da(Eu) = (1.0-2.1)·10-15 m2/s. With cobalt and cesium, classical in-diffusion experiments were also performed for comparison, and similar Da values were obtained but with a large dispersion. To analyze the possible effects of the paper filter impregnated with the tracer on the determinations of Da with the analytical solution, one experiment was also analyzed using a detailed stochastic model of the setup. The good agreement between the two modeling approaches confirms the validity of this experimental setup and the analytical model fitting procedure.

A combination of two nuclear ion beam techniques, Rutherford Backscattering Spectrometry (RBS) and micro-Particle Induced X-Ray Emission (uPIXE) was tested to evaluate both diffusion profiles and radionuclide spatial distribution of radionuclides (RN) onto cement-based materials. The methodology was tested on a Spanish reference backfill concrete, used as engineering barrier in low-level radioactive waste repositories, using two elements (Cs and I) with different sorption behaviour onto the cement. The applicability and limitations of the selected methodology is discussed for both elements.