Field experiments have run for over 14 years to evaluate the behaviour of the same high-sodium content radioactive waste borosilicate glass buried in a loamy soil (glass K-26) and in an open testing area (glass Bs-10). Processing of field data for glass Bs-10 tested in an open area has resulted in a dissolution rate r = 0.42 µm/y and caesium diffusion coefficient D ≍ 1.8 10−20 m2/s at testing temperatures up to 19 oC. Both ion-exchange and hydrolysis control the corrosion of this glass. Processing of field data for K-26 glass revealed an insignificant role of glass dissolution. The caesium diffusion coefficient was estimated as D ≍ (3.4-5.1) 10−21 m2/s. Due to the relatively low storage temperatures (4.5 oC) used the leaching behaviour of glass K-26 is believed to be controlled by ion exchange processes. This mechanism is likely to remain dominant until the decay of 137Cs in the glass is below exemption levels.

The DOE is currently preparing a license application for the permanent disposal of high level radioactive waste at Yucca Mountain, Nevada. The proposed design of waste packages for the disposal of high level radioactive waste consists of an outer container made of Alloy 22, a corrosionresistant Ni-Cr-Mo-W alloy, surrounding an inner container made of Type 316 nuclear grade stainless steel. Under conditions where passivity is maintained, the uniform corrosion rate of Alloy 22 is slow and long waste package lifetimes are projected. However, the initiation of localized corrosion such as pitting or crevice corrosion may decrease waste package lifetimes. In this study the crevice corrosion susceptibility of Alloy22 was determined in chloride solutions with additions of oxyanions that are present in the groundwater at the potential repository site. When present in sufficient concentrations relative to chloride, nitrate, carbonate, bicarbonate, and sulfate inhibited pitting and crevice corrosion of Alloy 22.

Hollandite, based on Ba and Ti, has been shown to be an ideal medium by which active 135Cs and 137Cs can be immobilised. Diffraction measurements have shown that Cs modifies the symmetry of the material, from monoclinic to tetragonal. Experiments carried out here show the effects of Cs on two hollandite systems based on Ba-Al-Ti and Ba-Mg-Ti. Structural measurements are reported for neutron diffraction, and MAS NMR. The results show that Cs modifies the overall crystal symmetry while having little effect on the order/disorder of Al-Ti and Mg-Ti octahedra.

We model preferential release of 237Np, 234U, 230Th, 226Ra, and 210Pb from disposed commercial spent nuclear fuel as a result of alpha recoil damage, using the U.S. Nuclear Regulatory Commission (NRC) Total-system Performance Assessment (TPA) model for the potential repository at Yucca Mountain. Time-dependent augmentation of the ingrown component is simulated by increasing the initial parent inventory; we have used a factor of five increase, based on natural system observations. For 237Np, the magnitude of preferential release is subject to solubility limits. Stochastic TPA runs show a significant effect on modeled dose of preferential 237Np release, but low impact from the other four radionuclides. The mechanism could be ineffective if 237Np is incorporated into secondary phases. While our results are exploratory in nature, this approach to modeling decay-related enhancement of release can be applied in other nuclear waste disposal settings.

Apparent diffusivities (Da) for I− and Cs+ in compacted Na-smectite were studied as a function of smectite's dry density (0.9–1.4 Mg/m3, ionic strength (IS: [NaCl]=0.01, 0.51 M), temperature (22–60 °C) and diffusion directionto the orientated direction of smectite particles. The Da-values for both ions in parallel direction to the orientated direction showed a tendency to be higher than those in the perpendicular direction at lowIS. The Da-values for I− showed different trends depending on diffusion direction and dry density at high IS, but Da-values for Cs+ increased with increasing IS in both diffusion directions. The activation enthalpies (ΔEa) for I−, slightly higher than that of diffusivity in free water (Do) at low IS, similar level to that of Do at high IS, increased with increasing dry density. While, ΔEa-values for Cs+, clearly higher than that of Do for all conditions, increased with increasing dry density. Since it is known that interlayer distance depends on both dry density and IS, diffusive pathway is considered to differ depending on the charge of diffusion species.

Abstract:Thick section copper canisters are planned to be used as a corrosion protection of nuclear waste disposal containers for long term underground deposal in Sweden. The copper canisters will have the top and possibly the bottom lid welded to the canister walls using electron beam or friction stir welding. Due to the high external hydrostatic pressure and the relatively high temperature of the waste during the first one hundred years the copper will creep. The creep process will close the manufacturing gap between the cast iron container and the copper canister. The creep ductility must be sufficient to avoid cracking of the weld.

Specimens cut from the friction stir welds and the electron beam welds have been creep tested at temperatures ranging from 75 to 175 °C. Cross-weld specimens were used for both friction stir and electron beam welds. Weld metal, heat affected zone and base metal were also studied for friction stir welds. The results for the electron beam welds show that the main creep deformation is concentrated to the weld metal where the failure takes place. Weld metal and most cross-weld tests of friction stir weld material show similar creep lives and ductility as base metal tests. Ductility at rupture was found to exceed 30% for friction stir weld specimens, and the Norton power law exponent was determined to be between 30 and 50.

Abstract:Reaction between samarium trichloride and oxide ions was studied in the fused LiCl – KCl eutectic mixture at 723 K, by potentiometry with an yttria-stabilised zirconia membrane electrode. Titration curve demonstrated the existence of a precipitated samarium oxychloride (SmOCl), under this experimental conditions, with a solubility product value of 7.59 ţ 0.3 (molality scale). The combination of this value with the standard potentials allowed setting up the potential–pO2- stability diagram which summarises the properties of Sm species in the eutectic LiCl – KCl melt.

To investigate the behavior of monazite during accelerated radiation damage, which simulates effects of long term storage, 238Pu-doped polycrystalline samples of (La,Pu)PO4 and PuPO4 were synthesized for the first time ever and studied using powder X-ray diffraction (XRD) analysis and optical microscopy. The starting precursor materials were obtained by precipitation of La and (or) Pu from their aqueous nitrate solutions followed by calcination in air at 700°C for 1 hour, cold pressing, and sintering in air at 1200-1250°C for 2 hours. The 238Pu contents in ceramic samples measured using gamma spectrometry were (in wt.% el.): 8.1 for (La,Pu)PO4 and 7.2 for PuPO4. The (La,Pu)PO4 monazite remained crystalline at ambient temperature up to a cumulative dose of 1.19 × 1025 alpha decays/m3. In contrast, the PuPO4 monazite became nearly completely amorphous at a relatively low dose of 4.2 × 1024 alpha decays/m3. Swelling and crack formation due to the alpha decay damage was observed in the PuPO4 ceramic. Also, under self-irradiation this sample completely changed color from initial deep blue to black. The (La,Pu)PO4 monazite was characterized by a similar change in color from initial light blue to gray, however, no swelling or crack formation have so far been observed. The results of this study allow us to conclude that the radiation damage behavior of monazite strictly depends on the chemical composition. The justification of monazite-based ceramics as actinide waste forms requires additional investigation.

In the geological disposal system, natural barrier contains many selective flow-paths. Since cement used for the repository construction alters the condition of groundwater to a highly alkaline pH of about 13, such hyperalkaline plume would affect permeabilities of the flow-paths. To obtain more reliable estimate on the migration of radionuclides released from the repository, we must consider the changes in flow-paths with time and/or in space.

In this study, the influence of highly alkaline plume on the permeability has been examined, considering also the direction of flow. In order to simulate the flow-paths, the amorphous silica particles were packed in the column, and the NaOH solution (0.1 M) was injected continuously at a constant flow-rate into the column at room temperature. The change in the permeability was traced, and the concentration of silicic acid in the eluted solution was measured by using the silicomolybdenum-yellow method. It was confirmed that the difference of pH values at the inlet and outlet of the column was negligibly small (pH=13.0).

The experimental results showed that the change in fraction dissolved with time strongly depended on a flow-rate and a flow-direction. However, in the relation between the permeability and the fraction dissolved, the permeability did not change in the range of up to 0.35 in fraction dissolved. The SEM images of particle surface showed that the inner pores of particle increased in size. This suggested that, in this range of fraction dissolved, the porosity between particles is almost retained, while each particle dissolves mainly through its inner pores. Moreover, the dissolution rate in the column flow system was considered as being remarkably limited by diffusion process, in comparison with that estimated from the batch test.

We define four distinct thermohydrochemical environments for drip shield and waste package corrosion in the potential nuclear waste repository, referred to here as the Dry, Seepage + Evaporation, Seepage + Condensation + Evaporation, and the Seepage + Condensation environments. These environments are bounded by temperature and relative humidity conditions at drift wall and drip shield/waste package surfaces judged most likely to initiate fundamental changes in the quantity and/or chemistry of in-drift waters. The duration in which different environments might exist is evaluated by comparing simulated, time-dependent temperature and relative humidity curves for three different locations within repository drift 25. In-drift conditions and processes postulated to cause drip shield/waste package corrosion are evaluated within the context of the thermohydrochemical environments by various means, including analytical calculations and geochemical simulations. Of the four environments considered here, the Seepage + Evaporation environment presents the most significant potential for aqueous corrosion of drip shield and waste package materials, and may persist for approximately 500 years in center drift locations. The likelihood for corrosion in other thermohydrochemical environments is significantly lower, but may increase with the acquisition of new data or the demonstration of extenuating circumstances.

Dissolution experiments were performed with 238Pu-doped UO2 containing doping levels of ∼1–100 Ci/kg UO2, in 0.1M perchlorate with or without 0.1M carbonate, at pH 9.5, to investigate the effects of α dose-rate on UO2 dissolution rates and on Geff, the fraction of radiolytically produced H2O2 supporting dissolution. Dissolution rates increased by only about 10 times for a ∼100-fold increase in doping level in both solutions, but increased between 10 and 100 times in 0.1M carbonate, compared to dissolution rates in 0.1M perchlorate. Geff values increased with decreasing doping level, suggesting that the dissolution reaction is limited by the reaction rate between UO2 and H2O2 at high dose rates, but becomes limited by the rate of production of H2O2 at lower dose rates.

This paper examines the ability of electron energy-loss spectroscopy (EELS) combined with transmission electron microscopy (TEM) to detect both low concentrations of Np in a U matrix and to provide evidence for incorporation of Np in U(VI) phases. The case for U(VI) secondary minerals acting as solubility-controlling phases for Np in repository performance assessment models has not been fully established. Direct evidence for incorporation, rather than sorption, continues to be difficult to obtain. Detection of Np with TEM-EELS is hampered by the occurrence of a plural (multiple) scattering event from the more abundant U atoms (U-M5 + U- O4,5), that results in severe overlap on the Np-M5 edge at ∼3665 eV. By examining the energy âgap' between the Np-M5 and Np-M4 edges (184 eV), a method for observing Np independently of the plural scattering event has been established. Clear evidence of Np incorporation into synthetic studtite {(UO2)(O2)(H2O)2](H2O)2} and uranophane {Ca(UO2)2(SiO3OH)2(H2O)5} has been found with TEM-EELS. The EELS technique continues to remain an important tool for examining the potential for transuranic behavior in nuclear waste materials.

Numerous studies have shown that colloidal particles in groundwater can facilitate radionuclide transport in subsurface environments. A series of laboratory experiments was conducted to investigate the effects of radionuclide sorption onto colloids and the surfaces of rock fractures. This research especially focused on the kinetic behavior of the sorption process. A mixed solution of Cs and clay colloids was injected into a single artificial fracture in a granite column. Simulations were also performed to analyze the experimental results using a numerical code, COLFRAC, which describes colloid-facilitated solute transport in discretely-fractured media. The code allows for either equilibrium or kinetic sorption onto the colloidal particles. The experimental and analytical results indicate that transport of Cs is facilitated by the colloidal particles, which can sorb Cs and transport through the fracture. The analyses also illustrate the importance of evaluating parameters that describe kinetic sorption onto colloids. Furthermore, radionuclide transport is likely to be retarded as colloidal particles that sorb radionuclides are filtered on the fracture surface.

We have studied the aqueous durability of pyrochlore-structured yttrium-titanate (Y2Ti2O7) and Nd/Al-bearing zirconolite [(Ca0.8Nd0.2)Zr(Ti1.8Al0.2)O7] in both neutral and acidic solutions, with and without the presence of 0.001 M of NaF. Scanning Electron Microscopy (SEM), X-ray Diffraction (XRD), X-ray Photoelectron Spectroscopy (XPS) and Atomic Force Microscopy (AFM) were used to characterize the composition, structure and morphology of the pyrochlore (Y2Ti2O7) and zirconolite surfaces, both before and after static dissolution testing at 90 and 150°C for four weeks. The leachates were also analyzed using Inductively Coupled Plasma Mass Spectrometry (ICP-MS) to estimate the individual elemental releases. The results show that the presence of F- ions only has a significant effect in acidic media on the dissolution behavior of pyrochlore and zirconolite. This detrimental effect is more pronounced for pyrochlore than zirconolite; the Y2Ti2O7 surface was replaced completely by alteration products after dissolution testing at 150°C for 4 weeks in acidic media with 0.001 M fluoride ions.

Diffusion coefficients and accessible porosities for HTO were measured on 30 samples from the Andra Underground Rock Laboratory in Meuse/Haute-Marne (France) using the through diffusion technique. Two distinct geological formations were studied: Oxfordian limestone and Callovo-oxfordian argillite between 166 and 477 m depth. The experiences were carried out with synthetic pore-water from each formation. The measured values of the effective diffusion coefficients (De) are ranging from 2.6× 10-12 to 12.4 × 10-11 m2 s-1 while accessible porosities (ε) are between 2.4 and 24%. Good correlations are found between these two parameters.

Dissolution rates for spent fuel have typically been reported in terms of a rate normalized to the surface area of the specimen. Recent evidence has shown that neither the geometric surface area nor that measured with BET accurately predicts the effective surface area of spent fuel. Dissolution rates calculated from results obtained by flowthrough tests were reexamined comparing the cumulative releases and surface area normalized rates. While initial surface area is important for comparison of different rates, it appears that normalizing to the surface area introduces unnecessary uncertainty compared to using cumulative or fractional release rates. Discrepancies in past data analyses are mitigated using this alternative method.

Deaerated 5 M NaCl solution is alpha-irradiated up to 700 days in the presence of UO2 pellets. Experiments are conducted with 238Pu-doped pellets in pure brine and with undoped UO2 pellets in a 238Pu containing brine. The long-lived radiolysis products H2, O2 and ClO3- are formed in all cases proportional to the dose applied on the solution and with yields corresponding to 200, 80 and 7 nMol/(L*Gy).

The U concentrations increase with time but do not exceed about 10−4 Mol/L thus indicating Uranium to be solubility limited. Moreover, the rinse solutions of the vessels at the end of the experiments contain up to one order of magnitude more U than found in the solutions. The total amounts of mobilized U deviate by less than a factor 10 from each other regardless of the large ratio of surface dose rates (ratio up to 2300) or mean dose rates (ratio up to 100), applied while the reference experiment (UO2 pellet in pure brine) yields only some 10−7 Mol/L U which is over 3 orders of magnitude lower than observed in experiments with radiation present.

The simulation of the radiation chemical processes using a kinetic reaction model can reasonably reproduce the findings of the experiments with respect to the formation of H2 and O2 and give the order of magnitude for the concentrations of chlorine species and of oxidized UO2.

We report the initial results of modelling studies performed on the Dounreay Database of Spent Nuclear fuel particles recovered from the marine environment and coastal foreshore adjacent to the Dounreay Fast Reactor Complex located in the North East coast of Scotland. We demonstrate how interrogation of the fuel-particle frequency versus 137Cs isotopic distribution allows us to generate particle behaviour models and residence lifetimes of various particle subgroups. Similarly we find the recovered fuel particle's spatial distribution and shape characteristics are determined by environmental selection rules governed by properties of adjacent sediment and sediment transport mechanisms.

Analysis of the Spent Fuel particle database allows us to speculate on the original Site discharge pathways and fuel composition at the time of discharge. Further such studies correlate well with other sample morphology and compositional data and allow us to speculate upon the ultimate fate of different spent fuel types.

The transformation of UO2 into U3O8 is of technological and academical interest because of the severe consequences on the spent nuclear fuel management. The structural mechanism responsible for the isothermal transformation of UO2 into U3O8 seems still unclear. Several phases (UO2+x, U4O9, β-U3O7, α-U3O7, U3O8 were reported but their true structures, phase boundaries between their existence domains and matter transport processes are still a matter of debate. Gathering accurate information on the behaviour of uranium oxide is a key issue for understanding the behaviour of spent nuclear fuel. The chemical diffusion coefficient ( ~ D) of UO2+x was determined by electrical conductivity experiments. Measurements were performed in transient state for departure from stoichiometry in the range 0<x<0.17 (10-11<P(O2)<10-8 atm.)and for 973<T<1673 K. We have found that ~ D is a decreasing function of the departure from stoichiometry x. This behaviour was attributed to the presence of singly charged (2:2:2) Willis defects as suggested by equilibrium conductivity measurements. The decrease of Dchim can be explained by transport processes occurring via a dynamic exchange between isolated mobile defects and complex defects frozen in clusters or domains. At higher P(O2), near U4O9, the time to reach an equilibrium electrical conductivity value becomes increasingly long. This suggests the presence either of large defect aggregates or of complex defects arranged into domains. Furthermore, the analysis of the transport processes in non equilibrium conditions has allowed us to show that the results of ~ D are consistent with those of the oxygen diffusion coefficient within the P(O2) and temperature range of stability of the [2:2:2] clusters.

Chemical durability of fully radioactive, high-sodium borosilicate glass K-26 was evaluated using the product consistency test PCT-A. Examination revealed normalised leaching rates as high as 5.93•10−2, 4.05•10−2 and 2.93•10−2 g/m2•day for sodium, boron and silicon respectively. Data on chemical durability of glass K-26 are consistent with similar composition glasses. These are of particular interest for performance assessment models.