A model for acidity within pores within corrosion products on anodically-dissolving UO2 was developed using Comsol Multiphysics 3.2 to complement ongoing electrochemical measurements. It was determined that a depression of pH within pores can be maintained if: electrochemically measured dissolution currents used in the calculations are attenuated to reflect very localized pores; corrosion potentials exceed -250 mV (vs. SCE); and pore depths are > 1 μm for 300 mV or >100 μm for -50 mV (vs. SCE). Mixed diffusional-chemical equilibria control is suggested through deviations in the shapes between pH-potential and pH-pore depth plots.

The conditions of the bentonite in an engineered barrier for HLW disposal have been simulated in a laboratory test. Six cylindrical blocks of bentonite compacted at a dry density of 1.64 g/cm3 were piled up in a hermetic Teflon cell. The total length of clay inside the cell was 60 cm. The bottom surface of the bentonite was heated at 100°C while the top surface was injected with granitic water. The duration of the test was 7.6 years. The water intake was measured during the test and, at the end, the cell was dismounted and the dry density, water content, specific surface area, mineralogy, geochemistry and swelling capacity of the clay were measured in different sections along the blocks. The material used in this test is the FEBEX bentonite.

Based on a probabilistic model previously proposed, a Monte Carlo simulation code (EBSPA) has been developed to predict the lifetime of the engineered barriers system within the Yucca Mountain nuclear waste repository. The degradation modes considered in the EBSPA are general passive corrosion and hydrogen-induced cracking for the drip shield; and general passive corrosion, crevice corrosion and stress corrosion cracking for the waste package. Two scenarios have been simulated using the EBSPA code: (a) a conservative scenario for the conditions thought likely to prevail in the repository, and (b) a worst-case scenario in which the impact of the degradation processes is overstated.

During dry storage, the oxidation of the spent fuel in case of cladding and container failure (accidental scenario) could be detrimental for further handling of the spent fuel rod and for the safety of the facilities. Depending on whether the uranium dioxide is under the form of powder or pellet, irradiated or unirradiated, the weight gain curves do not present the same shape. To account for these different behaviours, two models have been developed. Firstly, the oxidation of unirradiated powders has been modelled based on the coexistence, during the oxidation, of two intermediate products, U4O9 and U3O7. The comparison between the calculation and the literature data is good in terms of weight gain curves and chemical diffusion coefficient of oxygen within the two phases. Secondly, the oxidation of spent fuel fragments is approached by a convolution procedure between a grain oxidation model and an empirical parameter which represents the linear oxidation speed of grain boundary or an average distance able to cover the entire spent fuel fragment. This procedure of calculation allows in one hand to account for the incubation period noticed on unirradiated pellets or spent fuel and in another hand to link the empirical parameter to physical as porosity, cracks or linear power, or operational parameters such as fission gas release (FGR) respectively. A comparison of this new modelling with experimental data will be proposed.

Concentrated NaCl solution was γ-irradiated in autoclaves under a pressure of 25 MPa. A set of experiments were conducted in 6 mol (kg H2O)-1 NaCl solution in presence of UO2(s) pellets; in a second set of experiments γ-radiolysis of the NaCl brine was studied without UO2(s). Hydrogen, oxygen and chlorate were formed as lonγ-lived radiolysis products. Due to the high external pressure, all radiolysis products remained dissolved. H2 and O2 reached steady state concentrations in the range of 5·10-3 to 6·10-2 mol (kg H2O)-1 corresponding to gas partial pres-sure of ∼2 to ∼20 MPa. Radiolytic formation of hydrogen and oxygen increases with the concentration of bromide added to solution. Both, in the presence of bromide, resulting in a relatively high radiolytic yield, and in the absence of bromide surfaces of the UO2(s) samples were oxi-dized, and concentration of dissolved uranium reached the solubility limit of schoepite / NaUO2O(OH)(cr) transition. At the end of the experiments the pellets were covered by a surface layer of a secondary solid phase having a composition close to Na2U2O7. The experimental results demonstrate that bromide counteracts an H2 inhibition effect on radiolysis gas production, even at a concentration ratio of [H2] / [Br-] > 100. The present observations are related to the competitive reactions of OH radicals with H2, Br- and Cl-. A similar competition of hydrogen and bromide, controlling the yield of γ-radiolysis products, is expected for solutions of lower Cl- concentration

The stability of spent nuclear fuel as a waste form largely depends upon the oxidation behavior of uranium dioxide (UO2) under repository conditions. Because U6+ phases are more soluble than U4+ phases, and changes in oxidation state can lead to volume changes which increase the susceptibility of spent fuel to corrosion, understanding the mechanisms behind the formation of more highly oxidized UO2 phases is important. While a variety of diffraction and spectroscopic studies have been used to investigate the oxidation behavior of UO2 to U4O9, questions still remain as to the effect of interstitial oxygen on the oxidation state of uranium in hyperstoichiometric UO2. In this study, computational techniques were used to investigate questions such as: Is the formation of U5+ or U6+ more energetically favorable upon the oxidation of UO2

A density functional theory (DFT) approach was used to investigate energetically favorable locations of interstitial oxygen atoms and to determine their effect on the charge distribution of uranium in U4O9, specifically. A single unit cell of U4O9 is used as a starting model, and the optimized geometry and charge distribution is tested by using starting models with different U charge assignments. Results from our calculations suggest that the formation of one U5+ per addition of interstitial oxygen at a perpendicular bisector site is favorable. Deflection of two lattice oxygen atoms along the body diagonal of the cubic sites is also observed upon the addition of one interstitial oxygen atom. Structural and bond length data are compared with experimental data whenever possible.

Passive and corrosion behaviors of Alloy 22 in chloride and fluoride containing solutions, changing the heat treatment of the alloy, the halide concentration and the pH of the solutions at 90°C, was investigated. The study was implemented using electrochemical techniques, which included open circuit potential monitoring over time, potentiodynamic polarization and electro-chemical impedance spectroscopy (EIS) measurements carried out at open circuit and at passiv-ity potentials. Corrosion rates obtained by EIS measurements after 24 h immersion in naturally aerated solutions were below 0.5 m/year. The corrosion rates were practically independent of solution pH, alloy heat treatment and halide ion nature and concentration. EIS low frequency resistance values increased with applied potential in the passive domain and with polarization time in pH 6 - 1M NaCl at 90ºC. This effect was attributed to an increase in the oxide film thick-ness and oxide film aging. High frequency capacitance measurements indicated that passive ox-ide on Alloy 22 presented a double n-type/p-type semiconductor behavior in the passive potential range.

Zirconolite is a candidate host material for conditioning minor tri- and tetra-valent actinides arising from enhanced nuclear spent fuel reprocessing and partitioning, in the case of disposal of the nuclear waste. Its chemical durability has been studied here under charged particle-induced radiolysis (He2+ and proton external beams) to identify the possible effects of water radiolysis on the dissolution rates in pure water and to describe the alteration mechanisms. Two experimental geometries have been used in order to evaluate the influence of the following parameters: solid irradiation, water radiolysis. In the first geometry the beam gets through the sample before stopping at the surface/water interface. In the second one the beam stops before the surface/water interface. Results on the elemental releases due to the enhanced dissolution of the zirconolite surface during charged particle-induced irradiation of water are presented. Under radiolysis, an increase of one order of magnitude is observed in the Ti, Zr and Nd elemental releases. No difference in the total elemental releases can be noticed when the solid is also irradiated.

Acoustic emission (AE) signals were continuously recorded and their intrinsic frequency characteristics examined in order to evaluate the mechanical performance of cementitious wasteform with encapsulated Al waste. The primary frequency in the power spectrum and its range of intensity for the detected acoustic waves were potentially related with appearance of different micromechanical events caused by Al corrosion within the encapsulating cement system. In addition the process of cement matrix hardening has been shown as a source of AE signals characterised with essentially higher primary frequency (above 2 MHz) compared with those due to Al corrosion development (below 40 kHz) and cement cracking (above 100 kHz).

The purpose of this paper is to report the results of a new model study critically examining the high temperature nuclear waste disposal concept at Yucca Mountain using MULTIFLUX, an integrated in-drift- and mountain-scale thermal-hydrologic model. In addition to new results the paper summarizes results of a previous study. The results show that a large amount of vapor flow into the drift is expected during the period of above-boiling temperatures in the emplacement drift. This phenomenon makes the emplacement drift a water/moisture attractor for thousands of years during the above-boiling temperature operation.

The evaporation of the percolation water into the drift gives rise to salt accumulation in the rock wall, especially in the crown of the drift for about 1500 years in the example. The deposited salts over the drift footprint, almost entirely present in the fractures, may enter the drift either by rock fall or by water drippage. During the high temperature operation mode the barometric pressure variation creates fluctuating relative humidity in the emplacement drift with a time period of approximately 10 days. Potentially wet and dry conditions and condensation on the surfaces over salt-laden drift wall sections are unfavorable to the storage environment. Corrosive salt accumulation during the above-boiling temperature operation must be sufficiently addressed to fully understand the waste package environment during the thermal period. Until the questions are resolved, a below-boiling repository design is favored where the Alloy-22 will be less susceptible to localized corrosion.

Precipitation processes driven in hydrothermal conditions were applied to the preparation of phosphate-based ceramics. Three systems composed by a crystallized precursor linked with a high temperature compound were particularly examined: M(OH)PO4 / M2O(PO4)2 (M = Th, U), MPO40.5 H2O / MPO4 (M = La - Dy) and Th2-x/2Anx/2(PO4)2(HPO4) H2O / β-Th4-xAnx(PO4)4P2O7 (M = U, Np, Pu). A significant improvement of several physico-chemical properties of the powders, especially in the sintering capability and the homogeneity of the final solids was evidenced when starting from the precursors. Furthermore, these phases were also found to control the solubility of lanthanides and actinides during leaching experiments when reaching the saturation conditions in the solution.

The overall goal of this project is to develop a model to predict microbial effects on the performance of a high-level radioactive waste (HLW) repository. As a first step, the effects of microbes on groundwater chemistry have been evaluated with the numerical code 'MINT', using data collected from the borehole HDB-6 in the Horonobe underground research laboratory (URL) in Japan. The MINT code models biochemistry and geochemical equilibrium, with consideration of transport of solute and microbial activity. The MINT code simulates the activities of 6 major groups of microbes, classified by their metabolism as 'aerobic', 'denitrifying', 'manganese reducing', 'iron reducing', 'sulfate reducing' and 'methanogenic'. The specific activity of each of these groups will depend on the redox potential (Eh) of the groundwater.

Sensitivity analyses were performed to investigate the consequences of changes in groundwater composition on the effects of microbial activity. This indicates that the activities of Sulfate Reducing Bacteria (SRB) and methanogens are relatively high. The concentration of dissolved methane produced by such microbial activity is seen to be influenced by sulfate concentration. Based on the observed data from Horonobe URL, the concentration in oxygen is relatively high and the activity of denitrifying bacteria is the highest of the major 6 groups of microbes. This can, however, be attributable to chemical / microbial contamination of the groundwater during sampling. The modeling results indicate that the concentration of dissolved oxygen and nitrate ion should be quickly reduced by microbial metabolism, reducing the redox potential to a level low enough for active methanogenesis to commence. Such assessment can be important to evaluate the reliability of sampling and measurement techniques for sensitive geochemical parameters in general - and microbiology in particular.

Cladding failure of fuel rods caused by hydride-induced embrittlement is a reliability concern for spent nuclear fuel after extended burnup. Uncertainties in the cladding temperature, cladding stress, oxide layer thickness, and the critical stress value for hydride reorientation preclude an assessment of the cladding failure risk. A set of micromechanical models for treating oxide cracking, blister cracking, delayed hydride cracking, and cladding fracture was developed and incorporated in a computer model. Results obtained from the model calculations indicate that at temperatures below a critical temperature of 318.5 °C [605.3 °F], the time to failure by delayed hydride cracking in Zr-2.5%Nb decreased with increasing cladding temperature. The overall goal of this project is to develop a probabilistic-micromechanical methodology for assessing the probability of hydride-induced failure in Zircaloy cladding and thereby establish performance criteria.

Canada's Nuclear Waste Management Organization has recommended to the Canadian federal government an adaptive phased management approach to the long-term management of used nuclear fuel. This approach includes isolation in a deep geologic repository. In such a repository, the fuel would be sealed inside a carbon steel-lined copper container. To assist the development of performance assessment models studies of fuel behaviour inside a failed waste container are underway. Using an iterative modeling and experimental approach, the important features and processes that determine fuel behaviour have been identified and studied. These features and processes are discussed and the results of studies to elucidate specific mechanisms and determine important parameter values summarized.

Rock–core column experiments were introduced to estimate the diffusion and sorption properties of Kuru Grey granite used in block–scale experiments. The objective was to examine the processes causing retention in solute transport through rock fractures, especially matrix diffusion used to estimate the importance of retention processes during transport in different scales and flow conditions. Rock–core columns were constructed from cores drilled into the fracture and were placed inside tubes to form flow channels in the 0.5 mm gap between the cores and the tube walls. Tracer experiments were performed using uranine, HTO, 36Cl, 131I, 22Na and 85Sr at flow rates of 1–50 µlmin-1. Rock matrix was characterized using 14C–PMMA method, scanning electron microscopy (SEM), energy dispersive X–ray micro analysis (EDX) and the B.E.T. method.

Solute mass flux through a column was modelled by applying the assumption of a linear velocity profile and a molecular diffusion. Coupling of the advection and diffusion processes was based on the model of generalised Taylor dispersion in the linear velocity profile. Experiments could be modelled applying a consistent parameterization and transport processes. The results provide evidence that it is possible to investigate matrix diffusion at the laboratory scale. The effects of matrix diffusion were demonstrated on the slightly–sorbing tracer breakthrough curves. Based on scoping calculations matrix diffusion begins to be clearly observable for non–sorbing tracer when the flow rate is 0.1 μl×min-1. The experimental results presented here cannot be transferred directly to the spatial and temporal scales that prevail in an underground repository. However, the knowledge and understanding of transport and retention processes gained from this study is transferable to different scales from laboratory to in–situ conditions.

A screening experiment was carried out where the organic phase consisted of different concentrations of 2,6-bis-(5,5,8,8-tetramethyl-5,6,7,8-tetrahydro-benzo[1,2,4]triazy-3-yl)-[2,2']bipyridinyl (CyMe4-BTBP) [1], which is the extracting molecule, dissolved in different diluents. The aqueous phase consisted of 0.01 M HNO3 and was spiked with trace amounts of 241Am and 152Eu acting as analogues for actinides and lanthanides. The ionic strength was kept constant at 1.0 M using NaNO3. Three of the selected diluents used to dissolve CyMe4-BTBP were abandoned: n-hexane, TPH and TBB. Another extraction experiment screened out anisole, 1,1,2,2-tetrachloroethane and benzaldehyde.

A kinetic experiment was then performed. Three very different kinetic behaviors were observed in three different organic systems. For a proper understanding of the kinetic mechanism, further investigations are needed.

Alkali-borosilicate glasses are widely used in nuclear industry as a matrix for immobilisation of hazardous radioactive wastes. Durability or corrosion resistance of these glasses is one of key parameters in waste storage and disposal safety. It is influenced by many factors such as composition of glass and surrounding media, temperature, time and so on. As these glasses contain radioactive elements most of their properties including corrosion resistance are also impacted by self-irradiation.

The effect of external gamma-irradiation on the short-term (up to 27 days) dissolution of waste borosilicate glasses at moderate temperatures (30¢X to 60¢XC) was studied. The glasses studied were Magnox Waste glass used for immobilisation of HLW in UK, and K-26 glass used in Russia for ILW immobilisation. Glass samples were irradiated under ×-source (Co-60) up to doses 1 and 11 MGy. Normalised rates of elemental release and activation energy of release were measured for Na, Li, Ca, Mg, B, Si and Mo before and after irradiation. Irradiation up to 1 MGy results in increase of leaching rate of almost all elements from both MW and K-26 with the exception of Na release from MW glass. Further irradiation up to a dose of 11 MGy leads to the decrease of elemental release rates to nearly initial value. Another effect of irradiation is increase of activation energies of elemental release.

The Nuclear Decommissioning Authority (NDA) has taken over ownership of the majority of the UK's separated civil plutonium stocks, which are expected to exceed 100 metric tons by 2010. Studies to technically underpin options development for the disposition of these stocks, for example by immobilization or re-use as fuel, are being carried out by Nexia Solutions on behalf of NDA. Three classes of immobilization matrices have been selected for investigation by means of previous studies and stakeholder dialogue: ceramic or crystalline wasteforms, storage MOx, and vitreous or glass-based wasteforms. This paper describes the preliminary inactive experimental program for the vitrification option, with results from a wide range of glass compositions along with conclusions on their potential use for plutonium immobilization.

Following review, four glass systems were selected for preliminary investigation: borosilicate, lanthanide borosilicate, aluminosilicate and phosphate glasses. A broad survey of glass properties was completed in order to allow meaningful evaluation, e.g. glass formulation, waste loading, chemical durability, thermal properties, and viscosity. The program was divided into two parts, with silicate and phosphate glasses being investigated by Nexia Solutions and the Immobilisation Science Laboratory (ISL) at the University of Sheffield respectively.

Samples of natural coffinite (USiO4·nH2O) from Grants uranium region, New Mexico were investigated in order to understand the alteration process of coffinite under reducing and oxidizing conditions. Alteration of the primary coffinite under reducing conditions was promoted by organic acids and as a result secondary coffinite precipitated. Subsequently oxidizing fluids altered the coffinite, and (Na,K)-boltwoodite and jáchymovite precipitated with no rare earth elements. Based on the charge balance calculation we suggest that the amount of U6+ in the coffinite is less than 0.2 [apfu] and U6+ is accommodated in the structure via substitution: U4+ + Si4+ ⇔ U6+ + 2(OH)-. The high and variable analysis total of electron microprobe indicates that H2O is not an essential component in coffinite structure. The U-Pb ages of coffinite formation vary from 36.6-0 Ma suggesting that the coffinite has precipitated continuously in this period and organic matter can preserve reducing conditions even when oxidizing conditions dominate.

A salt solution (doped with Tc-99), that simulates the salt waste stream to be processed at the Saltstone Production Facility, was immobilized in grout waste forms with and without (1) ground granulated blast furnace slag and (2) pretreatment with iron salts. The degree of immobilization of Tc-99 was measured through monolithic and crushed grout leaching tests. Although Fe (+2) was shown to be effective in reducing Tc-99 to the +4 state, the strong reducing nature of the blast furnace slag present in the grout formulation dominated the reduction of Tc-99 in the cured grouts. An effective diffusion coefficient of 4.75 x 10-12 (Leach Index of 11.4) was measured using the ANSI/ANS-16.1 protocol. The leaching results show that, even in the presence of a concentrated salt solution, blast furnace slag can effectively reduce pertechnetate to the immobile +4 oxidation state. The measured diffusivity was introduced into a flow and transport model (PORFLOW) to calculate the release of Tc-99 from a Saltstone Vault as a function of hydraulic conductivity of the matrix.