Tailored ceramics are promising potential hosts for the immobilization of actinides [1]. In this study the potential of a range of different ceramic systems were investigated for the incorporation of waste plutonium, using Ce as a Pu surrogate. Durable actinide containing minerals exist in nature and provided excellent target phases for the titanate, zirconate, silicate and phosphate based formulations examined here [2]. The Ce single phase solid solution limits for each system were established and the processing parameters required to produce high quality ceramic specimens were optimised. Importantly, this was achieved within the constraints of a generic processing route suitable for fabrication of Pu bearing samples.

Multipurpose transport, aging, and disposal casks are needed for the management of spent nuclear fuel (SNF). Self-shielded cermet casks can outperform current SNF casks because of the superior properties of cermets, which consist of encapsulated hard ceramic particulates dispersed in a continuous ductile metal matrix to produce a strong high-integrity, high-thermal-conductivity cask.

A multiyear, multinational development and testing program has been developing cermet SNF casks made of steel, depleted uranium dioxide, and other materials. Because cermets are the traditional material of construction for armor, cermet casks can provide superior protection against assault. For disposal, cermet waste packages (WPs) with appropriate metals and ceramics can buffer the local geochemical environment to (1) slow degradation of SNF, (2) reduce water flow though the degraded WP, (3) sorb neptunium and other radionuclides that determine the ultimate radiation dose to the public from the repository, and (4) contribute to long-term nuclear criticality control. Finally, new cermet cask fabrication methods have been partly developed to manufacture the casks with the appropriate properties. The results of this work are summarized with references to the detailed reports.

Water samples (referred to as puddle water samples) were collected from the surfaces of a conveyor belt and plastic sheeting in the unventilated portion of the Enhanced Characterization of the Repository Block (ECRB) Cross Drift in 2003 and 2005 at Yucca Mountain, Nevada. The chemistry of these puddle water samples is very different than that of pore water samples from borehole cores in the same region of the Cross Drift or than seepage water samples collected from the Exploratory Studies Facility tunnel in 2005. The origin of the puddle water is condensation on surfaces of introduced materials and its chemistry is dominated by components of the introduced materials. Large CO2 concentrations may be indicative of localized chemical conditions induced by biologic activity.

A general concept for the disposal of spent fuel in clay formations is based on the multibarrier principle. In such concept, the barriers for radionuclides released into the environment are the clay host rock, the backfill, the canister overpack and the fuel itself. The innermost barrier is the dissolution of UO2 matrix, which is the main component of spent fuel. However the dissolution of UO2 upon groundwater contact depends strongly on the (geo)chemical constraints prevailing in the repository. In order to determine in how far the clay properties influence the dissolution of spent fuel, two different kinds of clay were considered: Ca-bentonite which presents an initial oxidizing environment, and Boom Clay which is characterized by its strong reducing capacity. The experiments were carried out with depleted UO2 in presence of either compacted dry Ca-bentonite with Boom Clay groundwater or compacted dry Boom Clay with Boom Clay groundwater. The leach tests were performed at 25°C in anoxic atmosphere (glove box under 0.4%CO2/99.6%Ar) for 2 years. The U concentrations were sampled during these 2 years, once every 6 months and the amount of U was determined in the clay after 2 years in order to determine the dissolution rate. After 2 years, an unexpected uranium concentration was found 50 times higher in the system Boom Clay with Boom Clay water (2.10-7 mol/L) than in the system Ca-bentonite with Boom Clay water (4.10-9 mol/L), maybe resulting from a larger colloidal fraction in the system Boom Clay with Boom Clay water. Final results are expected to allow the comparison of the U retention capacity of Ca- bentonite and Boom Clay in anoxic conditions with the U retention of Boom Clay found in reducing conditions.

Many types of ceramics have been put forward for immobilisation of high-level waste (HLW) from reprocessing of nuclear power plant fuel or weapons production. After describing some historical aspects of waste form research, the essential features of the chemical design and processing of these different ceramic types will be discussed briefly. Given acceptable laboratory and long-term predicted performance based on appropriately rigorous chemical design, the important processing parameters are mostly waste loading, waste throughput, footprint, offgas control/minimisation, and the need for secondary waste treatment. It is concluded that the “problem of high-level nuclear waste” is largely solved from a technical point of view, within the current regulatory framework, and that the main remaining question is which technical disposition method is optimum for a given waste.

Neptunium is a major contributor to the long-term radioactivity in a geologic repository for spent nuclear fuel (SNF) due to its long half-life (2.1 million years). The mobility of Np may be decreased by its incorporation into the U6+ phases that form during the corrosion of SNF. The ionic radii of Np5+ (0.089nm) and U6+ (0.087nm) are similar, as is their chemistry. Experimental studies have shown that Np can be incorporated into uranyl phases up to concentrations on the order of 100 ppm. The low concentration of Np in the uranyl phases complicates experimental detection and presents a significant challenge for determining the incorporation mechanism. Therefore, we have used quantum mechanical calculations to investigate incorporation mechanisms and evaluate the energetics of Np substituting for U.

CASTEP, a density functional theory based code that uses plane waves to model the behavior of the valence electrons and pseudo-potentials to model the behavior of the core and inner valence electrons, was used to calculate optimal H positions, relaxed geometry, and energy of different uranyl phases. The incorporation energy for Np in uranyl alteration phases was calculated for studtite, [(UO2)O2(H2O)2](H2O)2, and boltwoodite, HK(UO2)(SiO4)1.5(H2O). Studtite is the rare case of a stable, naturally-occurring peroxide mineral, in this case a uranyl hydroxyl peroxide that forms in the presence of H2O2 from the radiolysis of H2O. For studtite, two incorporation mechanisms were evaluated: (1) charge-balanced substitution of Np5+ and H+ for one U6+, and (2) direct substitution of Np6+ for U6+. For boltwoodite, the H atomic positions prior to Np incorporation were determined, as well as the Np incorporation mechanisms and the corresponding substitution energies. The preferential incorporation of Np into different structure types of U6+ minerals was also investigated. Quantum mechanical substitution energies have to be derived at Np concentrations higher than the ones found in experiments or expected in a repository in order to avoid the calculation of unit cells that are too large to be handled quantum mechanically. However, the quantum mechanical results are crucial for subsequent empirical force-field and Monte-Carlo simulations to determine the thermodynamically stable limit of Np incorporation into these uranyl phases.

The potential of plasma vitrification for the treatment of a simulant Plutonium Contaminated Material (PCM) was investigated. It was demonstrated that the PUO2 simulant, CeO2, could be vitrified in the amorphous calcium iron aluminosilicate component of the product slag with simultaneous destruction of the organic and polymer waste fractions. Product Consistency Tests conducted at 90ºC in deionised water and buffered pH 11 solution show the PCM slag product to be durable with respect to release of Ce.

Carbon steel is one of the candidate overpack materials for high-level waste disposal and is expected to assure complete containment of vitrified waste glass during an initial period of 1000 years in Japan. Carbon steel overpack will be corroded by consuming oxygen introduced by repository construction after closure of repository and then will keep the reducing environment in the vicinity of repository. The migration of iron corrosion products through the buffer material will affect migration of redox-sensitive radionuclides. Therefore the authors have carried out electromigration experiments with source of iron ions supplied by anode corrosion of iron coupons attached to compacted bentonite. Authors tried to use plutonium in this experimental configuration to obtain the knowledge of migration behavior of actinides. Authors also used cesium as reference. The concentrations of iron and sodium showed nearly complementary distributions. It is expected that iron ion could migrate as ferrous ion through the interlayer of montmorillonite replacing exchangeable sodium ions in the interlayer. Concentration profiles of plutonium in bentonite grew as time supplying electric potential as long as 168 h. Plutonium migrated from the iron anode toward cathode as deeper than 1 mm of the interior of bentonite even in 48 h, though plutonium could not diffuse 1 mm for 2 years. On the other hand, profiles of cesium seemed to be controlled by ordinary diffusion because of large diffusion coefficient of cesium in bentonite as large as 10$^{-12}$ m$^{2}$/s. Drift of the cesium profile by electric potential gradient could be observed clearly after 240 h at individual experiment for cesium. Apparent dispersion coefficients of plutonium were calculated from the profiles and were as large as 10$^{-13}$ m$^{2}$/s. Since plutonium migration was accelerated by electric potential, plutonium chemical species would have positive charge and were estimated as PuOH$^{2+}$ or PuCl$^{2+}$ by the thermodynamic calculation. Thus this experiment can provide a diffusion field for cations under a reducing condition with ferrous ions in water-saturated bentonite.

Phase composition of the murataite-based ceramics containing 10 wt.% lanthanum, cerium, neodymium, europium, gadolinium, yttrium, zirconium oxides was studied. The ceramics were prepared by melting of oxide mixtures in glass-carbon ampoules in air at ∼1500° C. They are composed of predominant murataite-type phases and minor extra phases: rutile, crichtonite, perovskite, ilmenite/pyrophanite, and zirconolite (in the Zr-bearing sample only). Three murataite-related phases with five- (5C), eight- (8C), and three-fold (3C) elementary fluorite unit cell are normally present in all the ceramics. These phases form core, intermediate zone, and rim of the murataite grains, respectively. They are predominant host phases for the rare earth elements whose concentrations are reduced in a row: M-5C>M-8C>M-3C. Appreciate fraction of La and Ce may enter the perovskite phase.

For the construction of the geological disposal system, the use of the cementitious material may affect the permeability change of the natural barrier around the repository. In practice, the construction requires a large amount of cement to maintain the mechanical stability of the repository system, particularly in Japan. Since such material alters the pH of ground water to highly alkaline region, also the permeability change of natural barrier is one of the notable factors for the performance assessments of geological disposal system.

In the high pH region, the solubility of silicic acid is very high compared to that in the natural pH (around 8). Therefore, the highly alkaline groundwater would dissolve and alter a part of rock surface. Usui et al. (2005) reported that the change of the pore structure due to the chemical reaction is the key factor to consider the change of the permeability. Moreover, such a change of the pore structure was considered as the result of the spatial heterogeneity of chemical composition. Since such spatial heterogeneity exists also in the sedimentary rocks consisted of crystalline minerals such as quartz and feldspar, we need to examine natural rock, in order to obtain more reliable understanding about the change of permeability induced by the highly alkaline groundwater (plume).

In this study, as crystalline mineral, silica sand was packed in the column, and the effect of dissolution induced by the highly alkaline plume on the permeability-change was examined. The silica sand particles mainly consist of SiO2 partially including Al2O3, CaO, Na2O3, and K2O. The volumetric flow rate and the pressure difference between the inlet and outlet of the column were measured, and the permeability was calculated. At the same time, the concentration of elements in the fluid was measured by the ICP-AES.

The experimental result showed that permeability decreased gradually, although the silica sand was continuously dissolved in the column. The behavior of the permeability is considered to be the result from the rearrangement of the particles. In the column test using the silica sand as packed mineral, the flow-path seems to be clogged by the rearrangement of the particles rather than the increase of the pore between the particles.

In the January 2006 State of the Union address, President Bush announced a new Advanced Energy Initiative, a significant part of which is the Global Nuclear Energy Initiative. Its details were described on February 6, 2006 by the U.S. Secretary of Energy. In summary, it has three parts: (1) a program to expand nuclear energy use domestically and in foreign countries to support economic growth while reducing the release of greenhouse gases such as carbon dioxide. (2) an expansion of the U.S. nuclear infrastructure that will lead to the recycling of spent fuel and a closed fuel cycle and, through transmutation, a reduction in the quantity and radiotoxicity of nuclear waste and its proliferation concerns, and (3) a partnership with other fuel cycle nations to support nuclear power in additional nations by providing small nuclear power plants and leased fuel with the provision that the resulting spent fuel would be returned by the lessee to the lessor. The final part would have the effect of stabilizing the number of fuel cycle countries with attendant non-proliferation value. Details will be given later in the paper.

The evolution of environmental conditions within the emplacement drifts of a potential high-level waste repository at Yucca Mountain, Nevada, may be influenced by several factors, including the temperature and relative humidity within the emplacement drifts and the composition of seepage water. The performance of the waste package and the drip shield may be affected by the evolution of the environmental conditions within the emplacement drifts. In this study, tests evaluated the evolution of environmental conditions on the waste package surfaces and in the surrounding host rock. The tests were designed to (i) simulate the conditions expected within the emplacement drifts; (ii) measure the changes in near-field chemistry; and (iii) determine environmental influence on the performance of the engineered barrier materials. Results of tests conducted in this study indicate the composition of salt deposits was consistent with the initial dilute water chemistry. Salts and possibly concentrated calcium chloride brines may be more aggressive than either neutral or alkaline brines.

The Deep Geological Repository (DGR) is currently the most accepted management option for the isolation of high level radioactive wastes. The DGR is based on a multibarrier system, which will limit releases of mobile radionuclides to the biosphere. In the design of the repository the spent fuel is encapsulated in canisters of carbon-steel. The space between the canister and the host geological formation will be filled with bentonite buffer clay. Under the prevailing conditions in a DGR, both localized and generalized corrosion phenomena are possible.

Corrosion of the canister will result in formation of solid and gaseous corrosion products, which can influence the behaviour of both the canister and the bentonite. Many studies have been carried out in order to improve the knowledge on the reactivity of these barriers. Most of them have focused on the mineralogical alteration of the bentonite as a function of temperature, time, iron/clay and liquid/rock ratio in batch conditions. The aim of this study is to provide experimental evidences, at repository conditions, on chemical and mineralogical changes during the canister-compacted bentonite interaction: determination of secondary minerals and their alteration reactions, the advance of the corrosion front in the compacted bentonite, and changes in porosity, permeability and cation exchange capacity.

Performance assessment models of the U. S. repository at Yucca Mountain, Nevada suggest that neptunium from spent nuclear fuel is a potentially important dose contributor. A scientific understanding of how the UO2 matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of Np is needed to predict the behavior of Np at the fuel surface during aqueous corrosion. Neptunium would most likely be transported as aqueous Np(V) species, but for this to occur it must first be oxidized from the Np(IV) state found within the parent spent nuclear fuel. In this paper we present synchrotron x-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within uranium oxide spent nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the Pu and Np in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the UO2 matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl (UO22+) mineral assemblage that is depleted in Np and Pu relative to the parent fuel. The transition from U(IV) in the fuel to a fully U(VI) character across the corrosion front is not sharp, but occurs over a transition zone of ∼ 50 micrometers. We find evidence of a thin (∼ 20 micrometer) layer that is enriched in Pu and Np within a predominantly U(IV) environment on the fuel side of the transition zone. These experimental observations are consistent with available data for the standard reduction potentials for NpO2+/Np4+ and UO22+/U4+ couples, which indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potential of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions.

Investigations on the behavior of the radioactive element technetium frequently use a stable isotope of rhenium as an analogue. This is justified by citing the elements' similar radii, major oxidation states of +7 and +4, and eH-pH diagrams. However, recent studies (e.g.,[1] have shown this analogy to be imperfect. Therefore, one goal of our study is to compare the behavior of these elements with emphasis on the adsorption of perrhenate and pertechnetate (the major forms of Re and Tc in natural waters) on mineral surfaces.

Quantum mechanical calculations were performed on the adsorption of these two anions on relaxed clusters of the well-characterized sulfide galena (PbS) and some other Fe and S-bearing materials. With these calculations, we gain insight into differences between the anions' adsorption behavior, including geometry, adsorption energies, and electronic structure. Differences between interactions on terraces and step edges, the effects of co-adsorbates such as Na+ and Cl-, and chloride complexation are also explored. The influence of water was calculated using homogeneous dielectric fluids.

As a complement to the calculations, batch sorption tests are in progress involving ReO4-/TcO4- solution in contact with Fe metal, 10% Fe-doped hydroxyapatite, goethite, hematite, magnetite, pyrite, galena, and sphalerite.

Argonne National Laboratory has developed an electrometallurgical process for conditioning spent sodium-bonded metallic reactor fuel from the Experimental Breeder Reactor II (EBR-II). One waste stream from this process consists of a metal waste form (MWF) whose baseline composition is stainless steel alloyed with 15 wt% Zr (SS-15Zr) and whose microstructure is a eutectic intergrowth of iron solid solutions and Fe-Zr-Cr-Ni intermetallics. This paper reports scanning electron microscope (SEM) observations of corrosion products formed during static immersion tests in which coupons of surrogate MWF containing 10 wt% U (SS-15Zr-10U) were immersed in solutions with nominal pH values of 3 and 4 and 1000 ppm added chloride for 70 days at 50 °C. Although the majority of the surface areas of the coupons appear unchanged, linear areas with localized corrosion products apparently consisting of porous materials overlying corrosion-product-filled channels formed on both coupons, cross-cutting phase boundaries in the original eutectic microstructures. Many of the linear areas intersected the sample edge at notches present before the tests or followed linear flaws visible in pre-test images. Compositions of corrosion products differed significantly from the bulk composition, and the maximum observed concentration of U exceeded that reported in actinide-bearing phases in uncorroded surrogate MWF samples with comparable concentrations of U.

In general, the susceptibility of Alloy 22 to suffer crevice corrosion is measured using the Cyclic Potentiodynamic Polarization (CPP) technique. This is a fast technique that gives rather accurate and reproducible values of repassivation potential (ER1) in most cases. In the fringes of susceptibility, when the environment is not highly aggressive, the values of repassivation potential using the CPP technique may not be highly reproducible, especially because the technique is fast and because transpassive corrosion may influence or mask the nucleation and propagation of crevice corrosion. To circumvent this, the repassivation potential of Alloy 22 was measured using a slower method that combines Potentiodynamic-Galvanostatic-Potentiostatic steps (called here the Tsujikawa-Hisamatsu Electrochemical or THE method). The THE method applies the charge to the specimen in a more controlled way, which may give more reproducible repassivation potential values, especially when the environment is not aggressive. The values of repassivation potential of Alloy 22 in sodium chloride plus potassium nitrate solutions were measured using the THE and CPP methods. Results show that both methods yield similar values of repassivation potential, especially under aggressive conditions.

Fluidized Bed Steam Reforming (FBSR) is being considered as a potential technology for the immobilization of a wide variety of high sodium low activity wastes (LAW) such as those existing at the Hanford site, at the Idaho National Laboratory (INL), and the Savannah River Site (SRS). The addition of clay, charcoal, and a catalyst as co-reactants with the waste denitrates the aqueous wastes and forms a granular mineral waste form that can subsequently be made into a monolith for disposal if necessary. The waste form produced is a multiphase mineral assemblage of Na-Al-Si (NAS) feldspathoid minerals with cage and ring structures and iron bearing spinel minerals. The mineralization occurs at moderate temperatures between 650-750°C in the presence of superheated steam. The cage and ring structured feldspathoid minerals atomically bond radionuclides like Tc-99 and Cs-137 and anions such as SO4, I, F, and Cl. The spinel minerals stabilize Resource Conservation and Recovery Act (RCRA) hazardous species such as Cr and Ni. Granular mineral waste forms were made from (1) a basic Hanford Envelope A low-activity waste (LAW) simulant and (2) an acidic INL simulant commonly referred to as sodium-bearing waste (SBW) in pilot scale facilities at the Science Applications International Corporation (SAIC) Science and Technology Applications Research (STAR) facility in Idaho Falls, ID. The FBSR waste forms were characterized and the durability tested via ASTM C1285 (Product Consistency Test), the Environmental Protection Agency (EPA) Toxic Characteristic Leaching Procedure (TCLP), and the Single Pass Flow Through (SPFT) test. The results of the SPFT testing and the activation energies for dissolution are discussed in this study.

The rapid release of activity when water firsts contacts the spent fuel surface in disposal will depend on the pellet microstructure at the arrival time of water in the container. Research performed on spent fuel evolution in a closed system has shown that the evolution of microstructure under disposal conditions should be governed by helium behavior with the cumulated α{decay damage. The evolution of fission gas bubble characteristics under repository conditions has been assessed. In UO2 fuels with a burnup of 47.5 GWd/t, the pressure of fission gas bubbles with the input of helium atoms should not reach the critical bubble pressure, thus micro-cracking in grains is not expected.

A synroc-D ceramic consisting mostly of spinel, hollandite, pyrochlore-structured CaUTi2O7, UO2, and Ti-rich regions shows promise for immobilisation of a HLW containing mainly Al and U, together with fission products. Ceramics with virtually zero porosities and waste loadings of 50-60 wt% on an oxide basis were prepared by cold crucible melting (CCM) at ∼1500°C, and also by subsolidus hot isostatic pressing (HIP) at 1100°C to prevent volatile losses. PCT leaching test values for Cs were < 13 g/L, with all other normalised elemental extractions being well below 1 g/L.