This paper describes a novel approach to modelling radionuclide transport through a heterogeneous geological medium. It is assumed that a statistical description of the heterogeneity can be formulated. The approach then essentially consists of three steps: (i) sampling of a large number of potential transport pathways (15‘000 in this example), covering the ranges of properties; (ii) grouping of these pathways in a manner which reflects their importance for radionuclide transport into a small number of “legs” with appropriately averaged properties; (iii) modelling geosphere transport processes in these legs. It is demonstrated that the number of legs required to represent conservatively, but without excessive conservatism, all the transport pathways is small (5 in this example). This result is important for safety assessment, where many different parameter variations may need to be considered and where computational efficiency is often a vital issue.

Safety assessment of potential gas release from the L/ILW requires information on two-phase flow behavior of the host rock. To determine the relevant formation properties for gas flow through the host rock, a combination of hydrotest and gas-injection sequences were designed and subsequently implemented in the Nagra field testing program at Wellenberg. The results of the design caculations indicate that the gas threshold pressure can be estimated from a gas injection test followed by a shut-in test using the log-log diagnostic plot of the gas injection sequence and the Homer plot of the recovery sequence. For a composite well-aquifer model, the gas threshold pressure of the outer zone, when estimated from the gas injection sequence, is increased, because water is displaced from the inner zone into the outer zone as gas is injected into the inner zone. This results in the overestimation the gas threshold pressure of the outer zone, and in turn, of the air-entry pressures in the corresponding capillary pressure models. The pressure response during the gas injection sequence in a composite well-aquifer model can be drastically different for different relative permeability models. For a Brooks-Corey model, the pressure indicates a distinct increase when gas flow starts into the outer zone. For a Grant model, the pressures can actually decrease during a constant-rate gas injection sequence. A decrease in interval pressure during gas injection was observed in two different field tests in the Palfris Formation at Wellenberg. In borehole WLB-SB4as, the gas injection test in interval VM16 was analyzed with the inverse code ITOUGH2 to determine the two-phase flow parameters for two different conceptual models. The simulations indicated that the particular pressure response, characterized by a sudden decrease during gas injection, could not be reproduced with a Brooks-Corey relative permeability model. However, assuming a Grant model, a sudden pressure decrease during the gas injection sequence could be simulated. The results indicate the importance in determining the hydraulic properties and the well-aquifer model from a combined analysis of hydro- and gas injection tests, in order to accurately determine the two-phase flow parameters.

The crystal chemistry and microstructures of saleeite (Mg(UO2PO4)2•10H2O) and metatorbernite (Cu(UO2PO4)2•8H2O), from Koongarra, Australia and Shinkolobwe, Congo, were examined by X-ray diffraction analysis, infrared spectroscopy (IR), scanning electron microscopy (SEM) equipped with energy dispersive X-ray analysis, transmission electron microscopy (TEM) and analytical electron microscopy. The uranyl phosphates consist of uranyl phosphate layers with cations and waters in the interlayers. The IR spectra of saleeite and metatorbernite show the presence of hydroxyls in the interlayers in addition to water molecules. The d002 spacings of the hydrated phases of saleeite and metatorbernite up to 300°C reveal that the uranyl phosphate layers themselves are quite stable in the temperature range although the interlayer water molecules are lost easily. The presence of a mixed phase of saleeite and metatorbernite is confirmed in the micrometer and nanometer scales. However, SEM and TEM examination suggest saleeite and metatorbemite generally grow separately, and rarely form solid solution or interstratification. The results imply that U is retained in uranyl phosphate minerals even when the temperature at around repositories increases, and that saleeite and metatorbernite precipitate independently from solution according to their solubilities even when Mg2+ and Cu2+ coexist in solution.

A new approach to identifying and characterising system responses of the EBS based on analysis of ensemble runs of a generic and flexible source term model is developed and applied. A discussion of issues relating to the use of this methodology as a complementary approach to existing scenario development/safety case definition is also included.

This paper gives an overview of the numerical models employed for the simulation of ground-water flow at the Wellenberg site, the runs carried out and the kind of products which were specified as results of the modelling calculations. With respect to more conventional modelling approaches, based mainly on fully deterministic and steady-state schemes, this new approach integrates both deterministic and stochastic aspects (hydraulic conductivity distribution, geometry of the fracture network, etc.) in three different groundwater models, accounting for different scale levels. Moreover, this new approach is as close as possible to a realistic representation of the in situ conditions, since it is based on plausible distributions of transmissivity and hydraulic conductivity at all scales and takes the transient nature of flow fully into account. This latter aspect is fundamental since, at Wellenberg, both the underground facilities and a natural head anomaly in the host rock cause pressure disturbance in the flow field. Finally, and because boundary conditions (especially topographic surface and overburden) vary over time, model predictions based on current topography and climatic conditions are of limited duration and hence a validity duration has to be defined.

Wellenberg is the proposed candidate site for LLW/ILW repository in Switzerland (Figure 1). The performance assessment of the planned repository calls for a thorough hydrogeological characterization of the site. The spatial distribution of the hydraulic conductivity within the host rock is a key input for the evaluation of the host rock performances. The spatially variable conductivity field is modelled geostatistically in 3 dimensions using a conditional simulation method. In order to account for the possible range of spatial variability, several simulations of the conductivity are generated. The conditional simulation of the conductivity - representing possible versions of the unknown reality - are used as input for numerical modelling. For each simulated conductivity field, the groundwater flow is numerically modelled using finite elements. This Monte Carlo approach allows the propagation of the input uncertainty due to spatial variability of the conductivity onto the predictions delivered by the flow model. With the help of this approach several specified performance criteria - together with their uncertainty - were estimated for the evaluation of storage capabilities of the Wellenberg host rock.

The approach followed by Nagra for the assessment of the hydraulic properties of the excavation disturbed zone (EDZ) around a potential future underground repository site for low-level and medium-level short-lived radioactive wastes is presented. In the approach particular emphasis was placed on assessing, under realistic-conservative assumptions1, an appropriate way to represent the EDZ within the context of performance assessment (PA). Calculated stress changes due to the openings were mapped onto changes of the mechanical aperture of the rock fractures. The resulting changes of hydraulic properties of the EDZ were then computed, with analytical (simple, continuum-type approximation) and numerical (fracture network representation) methods. The results of both computations were used as the basis to: i) provide hydraulic properties for the EDZ used in PA models; ii) assess whether these properties are conservative, i.e., higher values than expected, or not.

Batch sorption experiments were carried out to investigate the adsorption of humic acids (HA) and Eu-humate colloids onto repository components: Boom clay, bentonite and Fe2O3 as a function of pH. The concentration of humic acids is 150 mg/l and the bulk concentration of Eu is 10−6 M. Radioactive 152Eu was used as a tracer. The adsorption of humic acids onto bentonite and Fe2O3 decreased with increasing pH. Humic acids were probably complexed to Al/Fe sites on the adsorbents through COOH and phenolic OH groups. It is difficult to interpret the adsorption of humic acids onto Boom clay because Boom clay itself contains high content of organic matters. The adsorption of Eu-humate colloids onto repository components were performed near neutral pH conditions. Eu was strongly adsorbed by three adsorbents, and pH had little effect on the adsorption in the neutral pH range. The strong complexation of Eu with humic acids did not prevent Eu from being adsorbed. Eu is likely bound to solid surfaces via humic acids. The adsorption of Eu-humate colloids onto mixed adsorbents (Boom clay + bentonite, Boom clay + Fe2O3, bentonite + Fe2O3, and Boom clay + bentonite + Fe2O3) was investigated. Most of Eu-humate colloids was adsorbed by the mixed adsorbents. The desorption of humic acids and Eu from repository components was also examined.

Nearly impermeable host rocks have been recognized as favorable media for many kinds of underground utilization such as radioactive nuclear waste disposal, storage of oil and LP gas, and CAES. To properly evaluate the ability of a geologic medium to retard transmission of fluids, it is necessary to accurately measure its hydraulic properties, most notably the permeability and specific storage. This paper presents a new flow pump permeability test system capable of testing low-permeability rocks under high confining and high pore pressure conditions, which simulate ground pressures at large depths. The new system was used to test the Inada Granite from Japan. The results of present study show that: 1) both permeability and specific storage of the rock are dependent not only on the confining pressure but also on the pore pressure. They decrease with the increment of the effective confining pressure, i.e., the difference between confining and pore pressures; 2) the permeability and specific storage of Inada Granite range from 10−11 to 10−12 cm/s and 10−6 to 10−7 1/cm, respectively. The flow pump technique with its rigorous theoretical analysis can be used to effectively obtain such low permeabilities within several tens of hours; 3) the storage capacity of flow pump system itself decreases with the increment of fluid pressure within the permeating system.

The recent advances in HLW partitioning indicate that cesium–137 and strontium–90 can be separated in the form of a pure concentrate. Based on this concentrate the glass RT2–1 containing 4.7 wt.% Cs2O and 2.7 wt.% SrO was selected for more intensive and wider studies to predict long-term performance of this waste form under repository conditions.

Leach tests were made in distilled and mineralized water before and after crystallization at 20, 90 and 130°C.

Fully devitrified glass contained only one crystalline phase corresponding to the pollucite structure.

The first aim of the work was to calculate, by means of the computer codes ORIGEN2. 1, ITS 3.0 and TRIM95, the beta/gamma cumulative dose and displacements produced by the fission products hold in the high level wastes (HLW) from reprocessed CANDU spent fuel, up to one million years, with a burnup of 7,000 MWd/tU and a cooling time of 20 years. Based on that results were calculated and graphically presented the mean electron energy (β−, internal conversion and Auger) and the cumulative beta events and ionization dose due to electrons, per metric ton of heavy metal and per cubic centimeter of the sintered HLW glass form, as a function of the time up to one million year.

The scenario of spent fuel direct storage in a deep geological repository is currently under consideration in many countries. In this approach the nuclear fuel — mainly uranium dioxide —acts as a first barrier towards fission fragment and actinide release. Although the access of water in such a repository is rather unlikely, evaluating the safety of a disposal requires the knowledge of the mechanisms and kinetics of alteration and dissolution of the nuclear fuel in groundwaters. Despite the fact that the reported litterature on the topic is really extensive, it is still a challenging problem due to the complexity of the system: the used nuclear fuel cannot identify with uranium dioxide itself due to both radiation damage and modification of the matrix composition during reactor operation. Parametric studies in well-defined conditions on simple systems are therefore of prime importance in order to determine the relevant parameters involved in the structural transformations of the uranium-oxygen-water systenL The aim of the present work was to study the incidence of the electronic stopping of fission fragments on the behavior of nuclear fuel towards leaching in deionized water under oxidative conditions. Several features were of special interest: (i) quantification of the damage induced by swift-heavy ion irradiation prior to aqueous corrosion; (ii) stoichiometry and oxygen depth concentration of the alterated surface region; (iii) structural transformation of the crystalline structure induced by incorporation of additional O atoms; (iv) comparison between unirradiated and irradiated samples

Geologic disposal scenario combined with multiple-recycle P&B/T (partitioning, and burning and/or transmutation) treatment has some potential merits to complete the current geologic disposal scenario, and is discussed herein based on three plus two criteria for partitioning. The important gain of P&B/T treatment is that an unanticipated uncertainty due to geologic behavior for near and far field of repository can be avoided without verification for super-long period. The number reduction of GSC (glass solidified canister, i.e. waste package)is another important gain of P&T treatment to decrease the cost of geologic disposal, because the number reduction of GSC can save the total cost for the current disposal scenario

Cubic zirconia ZrO2 and zircon ZrSiO4 are considered as perspective crystalline form for immobilization of U, Pu and other radionuclides [1, 2]. The present study was undertaken to determine temperature dependence of solid solution limits in these compounds using thermodynamic theory of solid solutions and available equilibrium data. Thermodynamic mixing properties has been evaluated by means of the relationship between Margules parameter W and interatomic distances R in end-members AX and BX of a solid solution (A1−xBx)X

W = α(ΔR/R)2

where ΔR = R(BX) - R(AX), R = (1−x) R(AX) + x R(BX), α is a constant for isotypic isostructural compounds. According to the Urusov's energetic theory of enthalpy of mixing [3] such relationships exist between parameters W, AR and R of solid solutions with similar bonds between isomorphous atoms and their neighbors

We have conducted a series of experiments using laser spectroscopy and analytic electron microscopy (AEM) to study the microscopic effects of radiation damage in single crystals of LuPO4 doped with 244Cm3+ ions. The crystals (body-centered tetragonal, zircon-type structure) were grown in 1980 using a flux method with an initial 244Cm(t1/2 =18.1 y) concentration of ∼1% by weight. The dimensions of the single crystals of Cm3+:LuPO4 are about 0.5×1.0×4.0 mm3. After 17 years, the crystals are no longer transparent. Dark brown to gray straps along the crystal c-axis are observable under an optical microscope. Macroscopic fractures are not obvious on the natural surfaces of the a- and b-planes. However, due to alpha-particle emission of 244Cm, significant microscopic structural damage, including microscopic cavities, has accumulated in this end member of the xenotime minerals

In Germany certain quantities of spent MOX fuel will be disposed in a rock salt repository. In order to assess the safety of the repository the corrosion behavior of spent MOX fuel in salt solution must be understood. Since material properties of spent MOX fuel are different compared to these of spent LWR UO2 fuel, it is not possible to deduce from corrosion behavior of the latter to that one of spent MOX fuel. It is expected that the most significant differences will occur in the dissolution kinetics, whereas similar solid alteration products are expected from the similar overall composition of the fuels. Various experiments were performed focused on the effect of fuel type on the dissolution kinetics. The results of this work are summarized and evaluated

Radionuclide release rates of HLW glasses in brines are normally orders of magnitude lower than glass corrosion rates. Various mechanisms were invoked to explain the experimental release; they may be controlled by the glass dissolution rate, solubility, sorption, coprecipitation, etc.. Glass matrix corrosion kinetics is described by a first order equation, where pH, S/V ratio, an initial constant rate and affinity term based on orthosilic acid activity are the key parameters. The dissolution rate decreases significantly as dissolved silica accumulates in solution

The direct disposal of spent fuel is one of the options considered in the Swiss high level waste management program. One of the important questions, within this program, is the heat generation from high-burnup UO2and MOX spent fuels. Depending on the repository boundary conditions (e.g. ambient temperatures at depth, thermal properties of the host rock etc.), on the maximum temperatures allowed in the near field and on the heat output of the fuel, it may not always be possible to completely fill the conceptual waste canister. The aim of this paper is to address the potential loading of spent fuel into canisters for different possible repository heat loading restrictions

Synroc is a candidate waste form for the immobilisation of high level radioactive waste (HLW)[1]. It is polyphase titanate ceramic principally comprised of zirconolite, hollandite perovskite and rutile (nominally CaZrTi2O7, (BaxCsy)[(Ti3+, Al)2x+y(Ti4+)8−2x−y]O16), CaTiO3 and TiO2 respectively). Waste species substitute into the three former phases. In particular, actinides (ACTs) substitute onto the Ca and Zr sites in zirconolite and the Ca site in perovskite. Consequently over time, these phases will suffer alpha-recoil and alpha particle damage while hollandite and rutile will suffer alpha particle damage. The effect of radiation damage on the structure and consequently on the durability of Synroc's constituent phases is important to predictive modelling of Synroc's behaviour in the repository environment and risk assessment.

Zirconolite (CaZrTi2O7), pyrochlore (vIIIA2VIB2IVx6y) and perovskite (CaTiO3) are candidate phases for the immobilisation of rare earth elements (REEs) and actinides (ACTs) in various high level radioactive waste (HLW) forms [1]. The effect of radiation damage on the structure and consequently on the durability of these phases is important to predictive modelling of their behaviour in the repository environment and risk assessment