The migration of radionuclides through rocks is one of the important paths to estimate the impact of radioactive waste disposal on the environment. Because some amount of the radionuclides is sorbed on the minerals of rocks during the migration, Kd model is considered to be used to estimate the radionuclide migration. However, long-term applicability of Kd model is not known.

The joint Nagra/PNC Radionuclide Migration Programme has been running for over ten years in Nagra‘s Grimsel Test Site in the central Swiss Alps. The programme is specifically aimed at the further development of conceptual models of radionuclide transport in the geosphere, rigorously testing the applicability of current transport codes to quantify radionuclide migration in situ and assessing how successfully laboratory sorption data (specifically, Kd values) may be extrapolated to in situ conditions to predict radionuclide retardation in the geosphere [1]. A large series of field tracer migration experiments was carried out in a hydrologically well characterised water-bearing, complex fracture (or shear zone), increasing in complexity from simple, nonsorbing fluoresceine (a fluorescent dye), 3H, 3,4He, 82Br and 123I through weakly sorbing 22.24Na,85Sr and 86Rb to a final, long-term experiment with moderately sorbing 134,137Cs. The radionuclides were injected into a dipole flow field where the flowpath length, dipole width or shape and groundwater flow velocity were all varied. After a considerable learning period, generally good fits could be obtained between transport code predictions and subsequent field tracer breakthrough curves, suggesting that the transport codes tested were a reasonable representation of in situ conditions.

An attempt was made to analyze the mechanism of the distribution of Sr and Cs isotopes depending upon petrochemical and petrophysical characteristics of metavolcanites of the basement forming the territory of the Production Association (PA) Mayak, Chelyabinsk District, Russia. It was found that radionuclide reterdation depends on the structure, density and mineralogy of intact rock.

Natural fission reactors located in the uranium ore deposits of Oklo-Okélobondo and Bangombé (Gabon) are a unique case of natural nuclear waste disposal. The Reactor Zones (RZ) (up to fifteen) cores are made of 60% uranium (UO2) embedded in a clay matrix [1]. The uraninites are depleted in 235U and contain fission products. Therefore, they are considered to be good analogues for spent nuclear fuel matrix [2]. The aim of this work is to study the conditions of preservation over a long periode of time of the uraninites which have been sumitted to interraction with fluid having different oxidizing potential. For this purposes mineralogical (X-ray powder diffraction), chemical (electron microprobe) and oxygen isotopic analyses were performed on uraninites from various RZs.

Two billion years ago, natural fission reactions occured in the uranium ore deposit of Oklo (Gabon), mainly due to the high concentrations of fissile 235U. At that time, high uranium contents (40-60 % U) allowed reaching criticality when oxidised uranium bearing fluids met reduced fluids accompanying the migration of hydrocarbons. Natural fission reactors consist of a core made of uraninite crystals embedded in a clay matrix (mainly chlorites). The thickness of the clay layer range between few centimetres to 2 meters. Fission products are mostly located in the uraninite grains but newly crystallised minerals such as phosphate minerals, clays and metallic aggregates retain some proportions of migrated fissiogenic elements [1].

Geological formations are being considered as host media for nuclear waste disposal. The occurrence of natural U in rocks provides a possibility to test the radionuclide migration models used in safety studies of the disposal over comparable time periods. Here we study U accumulation into boulders as a process analogue for matrix diffusion; the boulders were found in glacial till in Hämeenlinna, southern Finland. Based on the glacial history of the site, matrix diffusion simulations, and independent U-series disequilibria (USD) dating, the U accumulation was interpreted to originate from the end stage of the latest glaciation, i.e. the system age is about 10 000 years1,2. The known time scale offers a rare opportunity for quantitative model testing; normally the time scale is difficult to determine for a single process in a natural analogue.

The U accumulation was earlier1,2 interpreted to be due to matrix diffusion and sorption. The postulated accumulation history consists of short in-diffusion and out-diffusion stages, as well as a longer chain decay stage. The in-diffusion was caused by U-rich waters discharging on the boulders at the end stage of the glaciation. The subsequent partial out-diffusion represents the period the boulders were temporarily submerged in the Yoldia sea during the early stage of the Baltic Sea. The final isolated radioactive chain decay stage began when the boulders, and their surroundings, rose above the sea level due to land uplift.

In this paper we report the first radiochemical results of a new larger boulder from the same area as the one studied earlier1; qualitatively, also the U distribution appears to be the same. Due to the larger dimensions, we can sample the inner zone of the boulder which matrix diffusion can not have reached within the postulated time, i.e. the state of the boulder before the U accumulation. The large amount of sample material containing almost only the recently accumulated U provides an opportunity to experimentally approach the kinetics of U fixation in situ. Understanding the long-term U fixation is essential in natural analogue studies, because the matrix diffusion model only has fast reversible adsorption (based on Kd) as the fixation process. Attempts to separate and quantify sorbed U in natural analogues have been reported elsewhere3.

Clay minerals are well known to be capable of retarding the movement of many potential pollutants, such as uranium, with their relatively high sorption capacities2. It has been suggested that high temperature thermal treatment processes are a potential method for the in situ immobilization of radionuclides or other contaminants2. The effect of progressive heating on the leaching behaviour of uraniferous clay material has therefore been investigated.

This paper discusses the results of post-closure safety studies for two different waste streams that, according to current Swiss waste-management concepts, may be co-disposed in a single deep geological repository. The waste streams are:

• directly disposed spent UO2 and mixed-oxide (MOX) fuel

• vitrified high-level waste from the reprocessing of spent fuel

The inventories are based on a consideration of the anticipated arisings from nuclear power generation in Switzerland. A part (at least) of these arisings will be reprocessed, with the resultant vitrified high-level waste and long-lived intermediate-level waste returned to Switzerland. The Swiss electricity utilities have placed contracts with BNFL (UK) and COGEMA (France) for the reprocessing of about one third of the total arisings of spent fuel from nuclear power generation, assuming an electricity production scenario of 120 GWa(e). The decision as to whether to reprocess the remainder is currently left open, implying that up to two thirds of the arisings could be in the form of unreprocessed spent fuel for direct disposal. For the purposes of the present study, however, it is assumed that the arisings from the full 120 GWa scenario will be either directly disposed or reprocessed.

Sorption behavior of Cs+ and Ba2+ ions on Magnesite was studied using the radiotracer method complemented by x-ray photoelectron spectroscopy (XPS). 137Cs and 133Ba were used as radiotracers. The batch method was used. Local tapwater containing primarily Na+, K+, Ca2+ and Mg2+ ions was used as the aqueous medium. For kinetic studies samples were shaken for periods ranging from one hour to seven days. Aliquots of the aqueous phase before and after shaking were counted using a 35 cm3 pure Ge detector. Magnesite samples were separated into various size fractions by dry and wet sieving. Both fractionated samples and unfractionated samples (particle size < 107mµ) were used. In the XPS the spectra were recorded using a KRATOS ES-300 spectrometer with Al Kα source. Sorption isotherm studies were carried out at 30, 40, 50, 60 and 70 °C.

Basaltic dykes in salt rocks can be used in two different ways as natural analogues for high level radioactive waste (HLW) in a salt repository: In the first approach the basalts serve as analogues for the corrosion behavior of HiLW glass products during millions of years because of many similarities in their physical and chemical properties.

A methodology was developed to determine the effectiveness of different components/barriers in isolating waste from the accessible environment for the potential nuclear waste repository at Yucca Mountain, Nevada [1]. Quantifying the performance offered by different barriers aids in prioritizing the testing and analysis activities aimed at resolving uncertainty in processes and parameters of greatest significance to long-term performance. If a substantial fraction of the overall performance of a repository system may be attributed to the performance of a particular barrier, then it is prudent to devote the necessary resources toward enhancing the confidence in our predictive ability regarding that barrier (through more in-depth characterization and testing).

The Wellenberg area, central Switzerland, is under investigation for a potential low and intermediate level radioactive waste repository. The host rock, the Palfris marl, is bounded on the north by Cretaceous limestone units of the Drusberg nappe and is underlain by a sequence of Tertiary to Jurassic sediments (limestones, marls) of the Wissberg-Firrenband-Equivalent. Chemical and isotopic data on groundwaters and rocks show that three essentially independent groundwater flow regimes occur in the area: (1) groundwaters of the Palfris marl, (2) groundwaters of the limestone units of the Drusberg nappe, and (3) groundwaters of the Wissberg-Firrenband Equivalent.

The preferred methods for nuclear waste disposal in Sweden are based on isolation in deep repositories in crystalline rock. As part of its research programme on the safety of final disposal, the Swedish Nuclear Power Inspectorate (SKI) initiated a project to examine how spatial variability in rock chemistry in combination with spatial variability in matrix diffusion affects the radionuclide migration along single fractures in crystalline rock.

A mathematical framework describing migration was developed with a numerical simulation package. In order to determine statistical patterns in geochemistry, in-diffusion, through-diffusion, and batch experiments will be performed. The purpose is to use the knowledge of statistical patterns in geochemistry as a basis for stochastic predictions which will be validated against the results of several laboratory migration experiments.

A medium grained granite and a diorite were selected at Äspö hard rock laboratory in southeastern Sweden. Four drill cores with a diameter of 20 [cm] have been collected from each of two rock types. One of the four drill cores in each series was sliced into cubes in order to evaluate sorption characteristics, porosity and effective diffusivity by in-diffusion and through-diffusion experiments on the individual pieces. The other three drill cores are to be used in three migration experiments. Experiments with the two rock types are run in two parallel series. Currently, all the laboratory experiments are underway.

The present paper describes the mathematical framework used for planning and interpreting experimental results. In particular, the formulations for sorption kinetics and matrix diffusion are crucial for distinguishing between the effects of various hydrodynamic and geochemical effects. Neretnieks [I] proposed a one-dimensional formulation for radionuclide migration that includes the effect of matrix diffusion and instantaneous matrix sorption. This concept has been further developed into two dimensional formulations [2], [3]. The model framework developed in this project [4] includes additional first order sorption kinetics in the rock matrix. Preliminary analyses indicate that sorption kinetics (non-equilibrium sorption) are often sufficiently pronounced so as to significantly affect interpretation of other phenomena affecting the migration process. Acknowledgement of sorption kinetics is, therefore, deemed important for reliable generalisations of the retardation of radionuclide migration.

Several investigations of Cs sorption onto minerals indicate that equilibration time varies from weeks in laboratory tests with illite and montmorillonite [5] to up to several years under special conditions for Chernobyl Cs in lake sediments [6]. Comans et al., [7], Nyffeler et al., [8] and Smith and Comans [9] discussed the different equilibrium times associated with the readily available binding sites on grain surfaces and less available sites such as frayed edges or in the grain interior. Skagius [10] conducted experiments with adsorption and desorption of Cs on crushed granite in different size fractions, the major constituents of which were quartz, feldspar and microcline. In some experiments the ratio between dissolved and particulate phases of Cs was still changing even after more than a years time.

In order to quantify kinetics of Cs adsorption on granite batch tests were conducted on crushed rock. Relationships between surface area available to sorption and transfer rate coefficients, as well as the ratio of dissolved to adsorbed mass phase species at equilibrium are established from the experiments. The specific surface area for the parent, intact rock, as well as coefficients of sorption kinetics, can be evaluated by combining the results with the crushed rock with those of the parent intact rock. Hence the effect of sorption kinetics on the solute pulse propagation along a single fracture can be numerically simulated. The maximum concentration of the breakthrough from a Dirac mass spike is higher than that obtained with equilibrium sorption. The kinetics of the adsorption process during the uptake phase can be interpreted as a decrease in the effective partition coefficient (adsorbed/dissolved ratio). In contrast, sorption kinetics during the release phase, can be interpreted as the results of an increase in the partition coefficient. This, in turn, causes a prolonged tail in the breakthrough curve. If a Heaviside step function is used as a boundary condition, simulation indicates an earlier arrival of the breakthrough curve front than that obtained with equilibrium sorption.

The impact of sorption kinetics on breakthrough curves is of the same order as the effect of shear dispersion. This is of great importance in the interpretation of the migration tests and to future generalisations to a field scale.

It is important to quantify the diffusion of radionuclides into pores in rock and to understand the diffusion mechanism to evaluate the migration of radionuclides in deep geological formations. The diffusivity of ions into the rock matrix is usually estimated using a theory for the diffusion in porous material, in which a proportional relationship exists between the effective diffusivity in rock and the diffusivity in the bulk of the solution. In this study, diffision of nonsorbing Γ and sorbing Sr2+ was observed simultaneously and the results were compared.

Six parallel, two-hour Workshops took place during the Symposium, covering subjects that were judged to be particularly important for waste management, directly relevant to the theme of the Symposium and of topical interest to numerous scientific groups in programmes around the world. The titles of the Workshop were:

The concept of such informal Workshops, intended to encourage open discussion, was welcomed by the Symposium participants. All Workshops were well-attended, with several attendees expressing the wish that the scheduling could have allowed participation in more than one group.

This workshop produced a lively discussion covering a number of issues relating to the transport of radionuclides through compacted bentonite. The predominant themes were:

– the structural and chemical inhomogeneity of the compacted bentonite barrier

– the process of radionuclide sorption in bentonite

– the postulated process of surface diffusion

The main points of interest in this discussion are summarised here and some general conclusions are drawn.

During the course of the title Symposium, six informal presentations were made. Structured discussion of research needs identified a priority list which is presented here.