The high level nuclear waste load is likely to increase in the future containment glasses, rising new questions about their long-term chemical durability. In this study sodium borosilicate compositions with Si substituted for RE (RE = rare earth) are altered under static conditions at fixed pH (8.5). The Na and B leached fractions reached just after the rate drop increase with the RE2O3 content. Part of the RE ions remain in the altered layer which is significantly de-polymerised (IR-ATR and NMR results). Their environment is reorganized and adopts a more centrosymetrical character. The higher leached fractions are attributed to rapid hydrolysis of Si/B-O-RE bonds with respect to Si-O-Si bonds.

When present, uranium is usually an element of importance in a nuclear waste repository. In the Waste Isolation Pilot Plant (WIPP), uranium is present in significant quantities, with about 647 metric tons to be placed in the repository [1]. Therefore, the chemistry of uranium, and especially its solubility, needs to be determined under WIPP-relevant conditions.

Long-term experiments were performed to measure the solubility of uranium (VI) in carbonate-free ERDA-6 brine, a simulated WIPP brine, at pCH+ values between 8 and 12.5. These data, obtained from the over-saturation approach, were the first WIPP repository-relevant data for the VI actinide oxidation state. The solubility trends observed pointed towards low uranium solubility in WIPP brine and a lack of amphotericity. At the expected pCH+ in the WIPP (˜ 9.5), measured uranium solubility approached 10-7 M. The objective of these experiments was to establish a baseline solubility to further investigate the effects of carbonate complexation on uranium solubility in WIPP brines, during the ongoing research program in actinide solubility under WIPP-relevant conditions.

137Cs and 90Sr, both byproducts of the fission process, make up the majority of high-level waste from nuclear power plants. 63Ni is a byproduct of the erosion-corrosion process of the reactor components in nuclear energy plants. The concentrations of these ions in solution determine the Waste Class (A,B, or C) and thus selective removal of these ions over large excesses of other ions is necessary to reduce waste and cut costs. Herein we report the use of the Inorganic Ion Specific Media (ISM) K2xMgxSn3-xS6 (x=0.5-0.9) (KMS-2) for the ion exchange of Cs+, Sr2+, and Ni2+ in several different conditions. We will also report the stability of this new material in the general conditions found at nuclear power plants (pH ˜6-8) and DOE sites (pH>10). Measurements at low concentrations were conducted with inductively coupled plasma mass spectrometry and Kd values are reported for each of the ions in a variety of conditions.

After a few hundred years, the actinides will dominate the radiotoxicity of spent nuclear fuel. This does not necessarily mean that the actinides will dominate the dose to organisms at the surface above a geologic repository. Quite the contrary, in most performance assessments this dose is dominated by long-lived fission products, activation products and, in the very long perspective, actinide daughters.

This makes the far-field migration properties of the actinides less interesting for further research. There are, however, other aspects of the presence of actinides in spent nuclear fuel and some of these and SKB's research in these fields is presented and discussed here.

The blocks of glassy material at 55 wt.% SB4 waste loading produced in a demountable cold crucible and cooled to room temperature in cold crucible and glasses cooled in a resistive furnace by a canister centerline cooling (CCC) regime were sectioned to investigate phase composition and elemental distribution between various parts of the blocks. X-ray diffraction (XRD), optical microscopy, scanning electron microscopy coupled with energy dispersive spectroscopy (SEM/EDS) and infrared (IR) spectroscopy studies revealed some difference in the texture but not in phase composition of the materials sampled from various parts of the blocks. The glass samples were composed of vitreous and spinel structure phases. Spinel was present as both skeleton-type aggregates of fine (micron- or submicron-sized) crystals segregated at early stages of melt solidification and larger (up to tens of microns) individual more regular crystals formed during slow melt cooling. There was some tendency for elemental segregation in the glass block from the cold crucible with enrichment of the deeper zones with heavier transition metal ions and depletion of Na, Cs, Ca, Al and Si. Uranium was quite uniformly distributed within zones of the block and entered the vitreous phase.

To identify the various oxalates and oxalato-nitrates likely to form during the nuclear fuel reprocessing we study crystallization of such compounds by various methods (slow diffusion, hydrothermal syntheses, in situ oxalate syntheses …), in different conditions and in presence of monovalent ions. In a first stage, lanthanides are used as surrogates of the actinides (III) radioactive elements. This communication reviews various lanthanides (III) compounds obtained by crystallization from nitric acid solution containing hydrazinium ions. Diethyl oxalate was used as a precursor for oxalate ions. A careful adjustment of the experimental conditions allowed us to synthesize single crystals of nitrates, oxalato-nitrates and oxalates with various ligand/Ln(III) ratio and containing nitrates as ligands or as counter ions. In all the compounds hydrazinium ions are present as counter ions. The crystal growth method is described and the crystal structures, determined by X-ray diffraction from single crystals, are discussed in terms of metal-oxalate frameworks.

Crystalline precipitates from molybdenum-containing nuclear waste glasses are complex, often containing multiple cations which confound routine structural techniques. A simplified mixed-alkali borosilicate model glass was found to have minor crystalline phases which could not be identified by x-ray diffraction. Multinuclear magnetic resonance (NMR) spectroscopy revealed sharp peaks characteristic of crystallinity superimposed on the broader glass signals, but were unattributable to any known molybdate phases. When a comprehensive range of cesium molybdates failed to reveal any matches with the observed 133Cs magic-angle spinning (MAS) NMR peaks in the composite glass/crystalline material, a series of mixed-alkali sodium-cesium molybdate phases was synthesized. 23Na, 133Cs and 95Mo MAS NMR revealed the formation of two mixed-cation molybdates which correlate with the observed NMR peaks for the phase-separated model glass. This work highlights the prominence of multiple crystalline phases in Mo-bearing nuclear waste glasses, and demonstrates the unique utility of solid-state NMR as a fingerprinting approach to identifying complex phases, especially where x-ray diffraction is limited by multiple phases, low concentrations or substitutionally disordered precipitates.

Thin crystals of La2O3, La2/3TiO3, La2TiO5, and La2Ti2O7 have been irradiated in situ using 1 MeV Kr2+ ions in the Intermediate Voltage Electron Microscope-Tandem User Facility (IVEM-Tandem), at the Argonne National Laboratory (ANL). We observed that La2O3 remained crystalline to a fluence greater than 3.1 × 1016 ions cm-2 at a temperature of 50 K. The four binary oxide compounds in the two systems were observed through the crystalline-amorphous transition as a function of ion fluence and temperature. Results from the ion irradiations give critical temperatures for amorphisation (Tc) of 840 K for La2Ti2O7, 865 K for La2/3TiO3, and 1027 K for La2TiO5. The Tc values observed in this study, together with previous data for TiO2, are discussed with reference to the phase diagrams for La2O3-TiO2 systems and the different local environments within the crystal structures. Results suggest an observable inverse correlation between Tc and melting temperature (Tm) in the two systems.

Fiber optic sensors offer a novel approach to monitoring of fractures in concrete waste disposal vaults and offer the possibility of determining the quantity, width and location of the cracks as they form. Fiber optics can directly detect cracks if they form within the path of a fiber optic as well as monitor secondary indicators of cracking such as temperature changes and strain. When cracks form in concrete waste disposal vaults they can fill with water which has a high heat capacity, this enables cracks to be observed by monitoring temperature variations near the crack. An analytical solution for heat transfer is applied to estimate the propagation of temperature waves around cracks. It is demonstrated that discharge rates through the concrete which are less than 10-5 m3/m-s do not produce a meaningful temperature wave through the concrete. Fractures in the concrete must be larger than 0.07 cm to produce a measurable result and temperature sensors must be located within 0.5 meters of a crack to detect a change in temperature produced by seasonal groundwater flow through a crack. A distributed system of fiber optic sensors may be embedded in the concrete vault and used to monitor crack formation, temperature variations and strain.

The behavior of two nuclear materials, namely cubic zirconia and urania, is investigated under different irradiation conditions in the low and medium ion energy range (tens of keV to a few MeV). In each case, these materials display a multi-step damage build-up, as revealed by both RBS/C and XRD measurements. It is demonstrated that each step exhibits characteristic features such as damage fraction, elastic strain, nature of defects, and thus presents a specific microstructure. The transition from one step to the following involves radiation defect re-organization which arises to lower the energy of the system.

Savannah River Site Defense Waste Processing Facility (DWPF) Sludge Batch 4 (SB4) high level waste (HLW) simulant at 55 wt % waste loading was produced in the demountable cold crucible and cooled to room temperature in the cold crucible. Appreciable losses of Cs, S and Cl took place during the melting. A second glass sample was subjected to canister centerline cooling (CCC) regime in an alumina crucible in a resistive furnace. X-ray diffraction (XRD) study showed that the glass blocks were composed of vitreous and spinel structure phases. No separate U-bearing phases were found.

Influence of operation factors in diffusion test of compacted bentonite (such as agitation of test solution in the reservoir, feed rate of the test solution and mass transfer resistance in the filter) on the diffusion data was examined by reservoir depletion (RD) test method using Cs+. The influence of these factors on the diffusion data was also analyzed based on the mathematical sorption-diffusion model which considered the feed of test solution and mass transfer resistance in the filter as well. The reservoir depletion data showed some remarkable influences of these operational conditions, especially in the system with low ionic strength. Change in mass transfer resistance at filter-compacted bentonite due to the operational conditions was found to be potential factor which disturb the diffusion data. The influence was reduced in the system with high ionic strength of solution.

The impact of Nd2O3, MoO3 and RuO2 addition on the competition between the crystallization of apatite Ca2Nd8(SiO4)6O2 and powellite CaMoO4 phases which both may appear in High Level Waste nuclear glass (under certain specific conditions of cooling and glass composition) has been studied on a simplified composition belonging to the system SiO2-Na2O-CaO-Al2O3-B2O3. X-ray diffraction (at room temperature and high temperature) and scanning electron microscopy measurements have been performed on five glasses under two different thermal treatments. We show that RuO2 acts as a nucleating agent for apatite. Moreover, neodymium and molybdenum cations seem to be very close in the glassy network as Nd2O3 addition stops the phase separation of molybdates and inhibits the crystallization of CaMoO4. On the contrary, MoO3 seems to favor the crystallization of apatite. For several samples, the evolution of the distribution of Nd3+ cations after crystallization was followed by optical absorption spectroscopy.

A three-dimensional (3-D) discrete fracture network (DFN) geo-descriptive model is developed for water conducting features (WCFs) in the sedimentary formations of Horonobe underground research laboratory (URL) in Japan. Fracturing and faulting system in/around the URL area, which is the main investigation area of the Horonobe URL project, is characterized by taking into account borehole geophysical logging data, regional geologic/structural data, and fracture/fault data (orientation, intensity, size) obtained from the surface-based investigations. Volumetric fracture intensity potential is estimated by the correlation and the multi-linear regression analysis of observed data, and is used as one of controls for 3-D DFN model. A regional scale 3-D geo-descriptive DFN model is constructed based on the analyzed fracturing system identified for the WCFs. The current 3-D geo-descriptive model could be utilized explicitly to derive performance assessment (PA) parameters for the hypothetical repository of the high-level radioactive wastes in Japan, and to assist optimization of the safe repository design.

The main indicators for the evaluation of the safety of a geological repository are the dose or the risk that are estimated for a sufficiently representative set of possible evolution scenarios. In recent years complementary sets of safety and performance indicators have been developed within national geological disposal programmes and international projects. Whereas safety indicators aim at giving an indication on the level of safety provided by the repository system, performance indicators aim at illustrating how the repository system works. Most sets of performance indicators that have been introduced hitherto in safety cases of geological repositories are related to the multi-barrier concept; they quantify or illustrate the contribution of the main engineered and natural barriers to the confinement of the radionuclides within the repository system. However, the application of the defence-in-depth principle to geological disposal has led to the introduction of safety functions in safety cases. Because of the paramount role played by the multi-safety-functions concept in recent safety cases, we have derived performance indicators that quantify the contribution of the main safety functions to the confinement of radionuclides in the geological repository system. The considered safety functions are containment, limitation of release and retardation. The proposed performance indicators are based on time-integrated activity or radiotoxicity fluxes released from the main successive compartments of the repository system. The proposed indicators can be applied to individual radionuclides as well as to a weighted sum of all radionuclides.

This paper describes the results from static leach tests using the ASTM International standard Materials Characterisation Centre (MCC-1) and Product Consistency Test (PCT) protocols for inactive High Level Waste (HLW) glasses fabricated at full scale on the Sellafield Vitrification Test Rig. The samples comprised monoliths and powders of a 75:25 Oxide:Magnox Blend glass with 31 wt% waste incorporation and a Magnox-only glass with 35 wt% waste incorporation. The tests were carried out in de-ionized water at 90 °C for durations up to 42 days and normalized mass losses calculated.

The results of MCC-1 and PCT tests on both 31 wt% Blend and 35 wt% Magnox glasses, showing measurable differences to the corresponding standard 25 wt% waste incorporation glasses, are presented. A series of Scanning Electron Microscopy (SEM) investigations were also undertaken. The variation in composition and thickness of the alteration layer with sample type and duration is reported.