A model has been developed to study the effects of secondary water radiolysis caused by dispersed radionuclides in a bentonite buffer surrounding a copper canister. The secondary radiolysis is the radiolysis caused by radionuclides that have been released from the spent fuel and are present either as solutes in the pore-water, as sorbed species on the surface of other minerals, or as secondary minerals. The canister is assumed to be initially defective with a hole of a few millimeters on its wall. The small hole will considerably restrict the transport of oxidants through the canister wall and the release of radionuclides to the outside of the canister. The dissolution of the spent fuel is assumed to be controlled by chemical kinetics at rates extrapolated from experimental studies. Two cases have been considered with the purpose to illustrate the behaviors of both conservative and non-conservative nuclides. The nuclides that are most relevant are those expected to be the dominant α-emitters in the long-term (e.g. 239Pu, 240Pu, 241Am). In the first case it is assumed that there is no precipitation of secondary minerals of the relevant radionuclides inside the canister. In the second case it is assumed that the radionuclide concentration within the canister is controlled by its respective solubility limit. The radionuclide released to the surrounding buffer is then predicted using a mass balance model. The modelling results show that in both cases, the spent fuel will not be oxidized at a rate significantly faster compared to the case where secondary radiolysis is completely neglected. In the first case, however, a large domain of the near-field can be oxidized due to a much faster depletion of reducing minerals in the buffer, compared to the case where secondary radiolysis is neglected. In the second case, the effects of the secondary water radiolysis will be quite limited.

Once groundwater intrudes into a damaged canister and wets the spent fuel pellets, radiation emitted from the spent nuclear fuel splits nearby water into oxidizing and reducing species. This may lead to an oxidizing condition near the fuel pellets. As a result, uranium oxide that makes up the fuel matrix will become more soluble, and the incorporated radionuclides will be released more rapidly. The dissolution process is, however, a dynamic one that can be influenced by many factors. Of great importance are the radiation power of the fuel matrix, the concentration of ligands near the fuel surface, and the transport resistance of the near field. Consequently, the escape of nuclides from the damaged canister is dominated mainly by the intrusion of ligands, and the precipitation/dissolution of secondary phases within the fuel rods. To investigate the possible effects of ligands and precipitates, a coupled dissolution and transport model, which includes the barrier effect of the Zircaloy claddings, is developed. The application of the model to a SKB-specified reference scenario indicates that by far the largest fraction of the oxidized uranium will reprecipitate within the canister. This may significantly decrease the fuel surface available for oxidation and the water available for radiolysis. Subsequently, much less fuel matrix will be dissolved and much less of the other nuclides will be released. Simulations further identify that carbonate and silicate have the greatest influences on the formation of secondary phases, and on the release of nuclides, under natural repository conditions.

In Italy a considerable amount of radioactive wastes has been produced during the about thirty years of nuclear activity. In 1996 began activities for the design of a LLW repository and for a site selection. A near surface LLW Repository based on vault concept is considered, for a national total inventory of around 100,000 cubic meters of conditioned wastes, mostly coming from dismantling of the phased out power stations. A near surface LLW Repository based on vault concept is considered.

The design activity under way includes the Engineered Barrier System (EBS) development and qualification. A cementitious material as backfilling barrier is being considered for waste isolation in a disposal module which would then be emplaced in the disposal unit. The construction material of the module is also under evaluation and qualification.

Possible interaction of the barrier with the chemical environment of the Italian waste inventory has also been identified. This paper describes some of the experimental activity under way for the mix-design identification for the cementitious barrier, along with the results of the first campaign of material characterization.

Different concretes and grouts were investigated with the main aim of assessing the long-term durability of these materials. Preliminary results indicate that crack-free and long-term durable concretes and grouts (300 years) can be manufactured even if exposed to aggressive environments. For the sake of brevity, in this paper the following characteristics only of the concretes will be considered:

- compressive, flexural and tensile strength

- carbonation tests

- chloride diffusion.

The GMT experiment (Figure 1) was initiated in the summer of 1997 under the auspices of RWMC (Radioactive Waste Management Center, Japan). It is part of the experiments conducted within Phase V (1997-2002) of Nagra's Underground Rock Laboratory at Grimsel, in the Swiss Alps [1]. The currents status and results from the site characterization activities and laboratory studies are presented. Future plans are also outlined.

FEBEX (Full-scale Engineered Barrier EXperiment) is a demonstration and research project dealing with the bentonite engineered barrier designed for sealing and containment of waste in a high level radioactive waste repository (HLWR). It includes two main experiments: an situ full-scale test performed at Grimsel (GTS) and a mock-up test operating since February 1997 at CIEMAT facilities in Madrid (Spain) [1,2,3]. One of the objectives of FEBEX is the development and testing of conceptual and numerical models for the thermal, hydrodynamic, and geochemical (THG) processes expected to take place in engineered clay barriers. A significant improvement in coupled THG modeling of the clay barrier has been achieved both in terms of a better understanding of THG processes and more sophisticated THG computer codes. The ability of these models to reproduce the observed THG patterns in a wide range of THG conditions enhances the confidence in their prediction capabilities. Numerical THG models of heating and hydration experiments performed on small-scale lab cells provide excellent results for temperatures, water inflow and final water content in the cells [3]. Calculated concentrations at the end of the experiments reproduce most of the patterns of measured data. In general, the fit of concentrations of dissolved species is better than that of exchanged cations. These models were later used to simulate the evolution of the large-scale experiments (in situ and mock-up). Some thermo-hydrodynamic hypotheses and bentonite parameters were slightly revised during TH calibration of the mock-up test. The results of the reference model reproduce simultaneously the observed water inflows and bentonite temperatures and relative humidities. Although the model is highly sensitive to one-at-a-time variations in model parameters, the possibility of parameter combinations leading to similar fits cannot be precluded. The TH model of the “in situ” test is based on the same bentonite TH parameters and assumptions as for the “mock-up” test. Granite parameters were slightly modified during the calibration process in order to reproduce the observed thermal and hydrodynamic evolution. The reference model captures properly relative humidities and temperatures in the bentonite [3]. It also reproduces the observed spatial distribution of water pressures and temperatures in the granite. Once calibrated the TH aspects of the model, predictions of the THG evolution of both tests were performed. Data from the dismantling of the in situ test, which is planned for the summer of 2001, will provide a unique opportunity to test and validate current THG models of the EBS.

The knowledge of pore water chemistry in the clay barrier is essential for performance assessment purposes in a nuclear waste repository, since the pore water composition controls the processes involved in the release and transport of the radionuclides. The methodology followed to define the representative composition of the FEBEX bentonite pore water is presented in this paper.

A series of bentonite-water interaction tests have been performed with the aim of providing a database on the main chemical parameters of the bentonite. These tests were carried out both with high solid to liquid (s:l) ratios (squeezing tests) and low s:l ratios (aqueous extracts tests). The exchangeable cations have also been analyzed to determine the selectivity coefficient of the exchange reactions. To complete the data set, a physical and mineralogical characterization of the bentonite was made.

The most significant bentonite-water interaction processes controlling the chemistry of the system was identified. The ion concentrations basically depend on the s:l ratio of the system, and the pore water composition is controlled by the dissolution of chlorides, dissolution/precipitation of carbonates and sulphates and the cation exchange reactions in the smectite.

The bentonite/water system was modelled with the PHREEQC2 program to obtain the best possible estimation of the pore water composition for initial conditions of water content (≍14%), after checking the conceptual model with the experimental results. The model predictions fitted satisfactorily with the experimental data at low s:l ratios. At high s:l ratios, the modelled results agree adequately, except for the sulphate content, which could be affected by the effective porosity, anion exclusion or stagnant zones not taken into account in the model. According to the model, the FEBEX bentonite pore water at 14% moisture is a sodium-chloride type, with an ionic strength of 0.25 M and pH of 7.78.

Na/K, Na/Mg, and Na/Ca exchange isotherms have been experimentally determined for the FEBEX bentonite. Na-homoionized FEBEXbentonite was reacted at room temperature with mixedsalt dissolution of NaCl/KCl, NaCl/MgCl2, or NaCl/CaCl2, while keeping a total cation normality of 0.5 eq L-1. Isotherm exchange experiments were performed using ten (duplicated)experimental points, which cover the complete range of the corresponding binary equivalent fractions. Results indicate that for the Na/K exchange reaction, Vanselow coefficients are larger than one, what is in agreement with the tendency of the smectite of having greater affinity for K than for Na. The exchange constant decreases as K progressively replaces Na in the smectite. This tendency ends when the equivalent fraction of potassium, EK, reaches a value of around 0.250.3. From this point to higher K contents, it remains nearly constant irrespective of EK but slightly decreasing again at values near one. The Vanselow selectivity coefficient for the Na/Mg isotherm indicates a preference for the divalent cation. It is nearly constant (Kv ≍ 5.6) for EMg < 0.6, but increases up to 10.2 for a nearly Mg-saturated smectite.

The Na/Ca exchange resembles that of Na/Mg, although the selectivity coefficients are larger (Kv ≍ 7.0 for ECa < 0.6). The comparison of the selectivity coefficients for the Na/Mg and Na/Ca exchange reactions indicates that the smectite has a slightly higher affinity for Ca than for Mg. This result is consistent with those observed for the Wyoming bentonite.

Ion exchange experiments have been performed with the FEBEX bentonite. Five grams of dry powder of this clay were put inside dialysis bags, which were located inside PFA reactors filled with 125 ml of water of a given chemical composition (Moody, Grimsel, and bentonitic-granitic type waters). The reactors containing the clay powder/water mixture were heated to different temperatures (from room temperature up to 80°C) along a time span ranging from 1 day to 1 year. Water was renewed according to a prescribed schedule but not the clay, which remained in place for the whole extent of each test. After each water renewal, major cations, silica, total inorganic carbon, and pH were analyzed. At the end of each test, the exchange complex and CEC of the bentonite were measured. These experiments have been modeled with a conveniently modified version of the EQ3/6 software package where ion exchange reactions were formulated as half reactions and added to its database. In general, model results are in fairly good agreement with experimental data, especially in the case of dissolved cations. Computed values of exchanged concentrations also match the measurements, although in some cases they deviate from them. The fact that the numerical results reproduce the observed patterns of exchange tests indicates that the adopted geochemical conceptual model is appropriate. Some features of the geochemical evolution of these tests also take place at the “mock-up” and “in situ” FEBEX tests.

Four kinds of diffusion experiments; (1) through-diffusion (T-D) experiments for diffusion direction dependency to compacted direction, (2) in-diffusion (I-D) experiments for composition dependency of silica sand in bentonite, (3) I-D experiments for initial bentonite grain size dependency, and (4) I-D experiments for the effect of a single fracture developed in bentonite, were carried out using tritiated water (HTO) to evaluate the effect of pore structural factors on diffusion. For (1), effective diffusivities (De) in Na-bentonites, Kunigel-V1Ŵ and Kunipia-FŴ, were measured for densities of 1.0 and 1.5 Mg.m-3 in the axial and perpendicular directions to compacted one. Although De values in Kunigel-V1Ŵ for both directions were similar over the density, De values for perpendicular direction to compacted one in Kunipia-FŴ were higher than those for the same direction as compacted one. For (2), apparent diffusivities (Da) in Kunigel-V1Ŵ with silica sand were measured for densities of 0.8 to 1.8 Mg.m-3. No significant effect of the mixture of silica sand was found. For (3), Da values for densities of 0.8 to 1.8 Mg.m-3 were measured for a granulated Na-bentonite, OT-9607Ŵ. However, no effect of initial bentonite grain size was found. For (4), Da values in Kunigel-V1Ŵ, in which a single fracture was artificially reproduced and immersed in distilled water, were measured. No effect of the fracture on Da was found. Based on this, it may be said that the composition of smectite in bentonite affects the orientation property of clay particle and also affects diffusion. Furthermore, a penetrated fracture formed in bentonite is restored for a short while and does not affect diffusion.

A fundamental concept combining a molecular behavior and macro-continuum phenomenon is presented for a multicomponent solute diffusion problem in bentonite clay, which is a key component of the Engineering Barrier System (EBS) of high-level radioactive waste disposal (HLRW). Bentonite is a micro-inhomogeneous material. Properties of the saturated bentonite are characterized by the constituent clay mineral (montmorillonite) and water, namely montmorillonite hydrate. We analyze its molecular behavior by applying a molecular dynamics (MD) simulation to inquire into the physicochemical properties such as diffusivity of chemical species. For extending the microscopic characteristics of constituent materials to a macroscopic diffusion behavior of the micro-inhomogeneous material we apply a homogenization analysis (HA).

Diffusion behavior of Ca2+ ions in water-saturated, compacted Na-montmorillonite was studied for the safety assessment of geological disposal of high-level radioactive waste. The diffusion coefficients of Ca2+ ions in compacted Na-montmorillonite obtained in this study were from 1.7×10-11 to 6.0×10-12 m2 s-1 with increasing dry density of 1.0 to 1.8 Mg m-3. These values were approximately a quarter those of Na+ ions at each dry density. This suggests that the diffusion of Ca2+ ions could be the rate-determining mechanism for the alteration of Na-montmorillonite into Ca-montmorillonite. The activation energy for diffusion of Ca2+ ions was almost equal to that in free water (17.3 kJ mol-1) at dry densities of 1.0 to 1.6 Mg m-3. However, the activation energy suddenly increased to 25.1 kJmol-1 at a dry density of 1.8 Mg m-3. This increase cannot be explained using a single-diffusion process model, such as the pore-water diffusion model. Diffusion behavior of Ca2+ ions in compacted montmorillonite along with the experimental data on the basal spacing and water content of montmorillonite, and the activation energy for diffusion of Ca2+ and several other ions are discussed.

Bentonite-water interaction was studied using a simple equilibrium model based on experimental measurements in order to describe bentonite porewater chemistry. Direct pH measurements for highly compacted bentonite and batch-type bentonite-water interaction experiments were performed under anaerobic conditions. In the direct pH measurements, resin particles doped with a pH indicator were sandwiched between a pair of bentonite columns immersed in a test solution. The experimental results showed that the solution compositions in equilibrium with bentonite depended on the bentonite to liquid ratio (B/L) and the initial solution composition. An equilibrium model assuming only fast equilibration processes between the bentonite minerals and the solution could be used to calculate the trends of pH and other ion concentrations with B/L. This study indicates that the surface deprotonation of smectite is a very important factor influencing the porewater chemistry in highly compacted bentonite.

The excavation damaged rock zone (EDZ) adjacent to the surface of deposition holes in a nuclear waste repository has been considered to be a potential pathway for the flow of water and the migration of radionuclides diffusing out of a waste canister via the bentonite barrier. The properties of the excavation-damaged rock zone adjacent to the surface of experimental deposition holes in the Research Tunnel at Olkiluoto on the southwest coast of Finland have been measured and are used in this study to evaluate the effect of the excavation damaged rock zone on the transport of radionuclides escaping from a waste canister. Since the hydraulic conductivity of compacted bentonite is low, the predominant mode of migration of nuclides through such material will be diffusion. The effect of the excavation-damaged rock zone on the transport of radionuclides was analyzed by comparing two different cases: 1) diffusion through the bentonite and 2) transport through the excavation damaged rock zone by the processes of advection and diffusion. According to this study, hydraulic gradients three orders of magnitude higher than those estimated to currently exist would be required to achieve the transfer of significant quantities of non-sorbing neutral nuclides through the excavation-damaged rock zone. According to this evaluation, the excavation damaged rock zone is unlikely to be a significant migration route in normal cases in the absence of large hydraulic gradients.

Microbial activity can have a significant impact on geochemical processes as it can influence mineral dissolution and precipitation, pH, alkalinity and redox. Experiments were conducted to study the rock-water and microbial interactions, specifically to examine the influence of microbes on the groundwater environment in fractured crystalline rock. These used combinations of crushed Äspö Diorite, Äspö groundwater with iron and sulphate reducing bacteria in columns and continuously stirred tank reactors under both anaerobic and aerobic conditions. Results showed loss (dissolution and/or mobilization) of fine grained crushed material (<50 μm) which had originally adhered to grain surfaces in the starting material. The mobilized fines were subsequently trapped on biofilms developed between grains. There was also evidence for the formation of smectite (not present in the starting materials) which, together with the trapping of translocated fines in intergranular pore throats by biofilms, resulted in blockage of the column experiments. The degree of mineralogical alteration and/or pore blocking was greater in the experiments when bacteria were present and occurred under anaerobic conditions. Little evidence for rock-water interactions was seen in the chemical analysis data for all the experiments. Modelling of the fluids showed them to be saturated with respect to clay minerals. The observed mineralogical changes seem to reflect alteration occurring in microbially mediated microenvironments close to mineral surfaces. Consequently, these changes may be too small to be detected by the chemical analysis of the bulk fluid. These experiments have shown that microbial activity can impact on fluid flow through porous media even in nutrient poor conditions. In addition, the formation of new clay minerals on pore-surfaces is potentially important for ion exchange and sorption reactions.

BNFL owns and operates the UK's principal solid Low Level Radioactive Waste disposal site at Drigg in Cumbria, north west England. Drigg has been receiving waste since 1959 with approximately 900,000 m3 of waste disposed of to date. Waste accepted for disposal at Drigg comes in a variety of forms including rubble, spoil, redundant equipment, scrap and process waste, and typically contains significant metallic and cellulosic components. The organic content of the waste means that microbial activity plays a significant role in the development of the repository environment. Consequently, microbial processes are integrated into many aspects of the Drigg Post-Closure Radiological Safety Assessment (PCRSA). This begins with the identification and screening of relevant features, events and processes, through supporting research, engineering designs and finally integration into radiological safety assessment modelling. This paper outlines how and where microbiology is integrated into the Drigg PCRSA and indicates areas of active research.

In order to assess microbial transport through compacted buffer material consisting of bentonite and silica sand with a density of 1.2, 1.6 or 1.8 g/cm3, sterilized test pieces with the diameter of 50 mm and the depth of 25 mm were fed with bacteria, Escherichia coli. The bacteria freely diffused when the proportion of Na-bentonite was 20 wt% or that of Ca-bentonite was 50 wt%. Bacterial movement was less than 5 mm in three weeks when the proportion of Na-bentonite was 70-100 wt% or that of Ca-bentonite was 100 wt%. A high proportion of bentonite is essential to slow down microbial entrance to the surface of a nuclear-waste container.

The lichen Trapelia involuta from uraniferous spoil heaps in Cornwall, England, growing directly on the secondary uranium minerals, metazeunerite and metatorbernite, was examined by field-emission scanning electron microscopy and transmission electron microscopy to assess the effect of Trapelia on uranium migration. We observed metazeunerite, sericite and scorodite as well as unidentified Fe-, Pb/As-, Fe/As-, Al/P-, Pb-bearing minerals concentrated in the lichen exciple and medulla. In addition, metazeunerite also occurred in the epithecium. The chemistries, sizes, and occurrences of the above minerals in the lichen suggest that fixation of U as well as Pb, As, Fe, and Al is dependent on lichen physiological processes. We suggest Trapelia accumulates these elements from groundwater and precipitates the above minerals within specific tissues. Our results indicate that some lichens retard uranium migration by accumulating uranium from groundwater and forming uranium-bearing minerals within their tissues.

A calorimetric investigation of the enthalpies of formation of Gd2(Ti2-xZrx)O7, where 0≤ × ≤ 2 is underway. All samples exhibit pyrochlore (Fd3m) peaks in their XRD patterns. However, where x=2 significant local disorder is observed in the Raman spectra. Preliminary data for the enthalpies of formation from the oxides in kJ/mol are: x=0, ΔHf = -113.4±2.7; x=0.5, ΔHf = -94.0±3.0; x=1.0, ΔHf = -74.2±4.9; x=1.5, ΔHf = -64.5±2.0; x=2, ΔHf = -52.2±4.8. Two additional samples, Gd1.80Zr2.15O7.00 (pyrochlore) and Gd2.15Zr1.87O7.00 (fluorite), were also studied. Their enthalpies of formation from the oxides in kJ/mole are -50.9±3.3 and -46.4±3.4 respectively. Replacing Ti with Zr, i.e. when x=2, destabilizes the pyrochlore in enthalpy by approximately 60 kJ/mol. The ΔHmix for the Gd2(Ti2-xZrx)O7 solid-solution series is positive and can be described by a regular solution formalism with an estimated interaction parameter, ŝ = +20 kJ/mol. The results of this study suggest that the pyrochlore to fluorite transition enthalpy in Gd2Zr2O7 is small, of the order of the configurational entropy contribution due to cation disorder at the transition temperature, TΔSconf. ≍ 10 kJ/mol.

In Japan, as outlined in the overall high-level radioactive waste (HLW) management program defined by the Japanese Atomic Energy Commission (AEC, 1994), HLW from reprocessing of spent nuclear fuel will be immobilized in a glass matrix and stored for a period of 30 to 50 years to allow cooling. It will then be disposed of in a deep geological formation. Pursuant to the overall HLW management program, an organization with responsibility for implementing HLW disposal will be established around the year 2000. This will be followed by site selection and characterization, demonstration of disposal technology, establishment of the necessary legal infrastructure, relevant licensing applications and repository construction, with the objective of starting repository operation by the 2030s and no later than the mid 2040s.

The HLW disposal program is currently in the research and development (R&D) phase and the Japan Nuclear Cycle Development Institute (JNC) has been assigned as the leading organization responsible for R&D activities. The aim of the R&D activities at the current stage is to provide a scientific and technical basis for the geological disposal of HLW in Japan and to promote understanding of the safety concept not only in the scientific and technical community but also by the general public. One of the features of the R&D program is that its progress is documented at appropriate intervals, with a view to clearly determining the level of achievement of the program and to promote understanding and acceptance of the geological disposal strategy by the general public. As a major milestone, the Power Reactor and Nuclear Fuel Development Corporation (PNC, now JNC) submitted a first progress report, referred to as H3 (PNC, 1992), in September 1992.

In 1992, the Commonwealth Government supported by States and Territories commenced an Australia-wide search for a site for a suitable site for a national radioactive waste repository for Australia's small inventory of low level and short-lived intermediate level waste. In 1998, the central-north region of South Australia was chosen as the preferred area in which to undertake detailed site investigations as it had the largest area of potential suitability compared to relevant technical and social criteria. Public consultation has been conducted both nationally and in the region.

Site investigations are currently in their final phase, and it is expected that a preferred site will be identified before the end of 2000. Environmental assessment and licensing processes will then commence, and the repository may be constructed in 2002.