The symposium “Scientific Basis for Nuclear Waste Management XVII” was held in June 15 – 19, 2003, in Kalmar, Sweden. The symposium, which was officially opened by the County Govenor of Kalmar County, Sven Lindgren, attracted 222 participants from 21 countries. Nearly 200 papers were presented during the three days of scientific sessions.

Organizing a symposium this size requires the assistance of a large number of people involved both in establishing the scientific program and in planning and executing the practical organizational arrangements. Our window to the world, the symposium's homepage, was continuously kept up to date through the excellent work of Jonny Rönnfjord.

Four nuclear power plant units have been operated in Finland over 20 years. The plants are located at two sites, Olkiluoto and Loviisa. Responsibility for nuclear waste management lies on the utilities, which have established a joint company POSIVA to take care of spent fuel disposal and other expert tasks of nuclear waste management.

Already in 1983 the Finnish Government set the objectives and the schedule for the national waste management programme. Since then, two shallow underground repositories have been constructed for low- and medium-level operating waste in crystalline rock at the power plant sites. At the end of 2002 the amount of operating waste emplaced in these two facilities was 4923 m3 (total accumulation 6724 m3).

Spent nuclear fuel is stored in interim storage pools at the nuclear power plant sites. The total accumulation was 1228 tU at the end of 2002. Today, the main activities in nuclear waste management concern spent fuel disposal deep in the Finnish bedrock. Apart from technical and scientific issues, a major challenge faced has been the creation of sufficient public acceptance, which is a must in order to obtain favourable political decisions.

The Swedish nuclear waste management programme has entered the site investigation phase. Early 2002 SKB received permission from the municipalities of Östhammar and Oskarshamn to perform site investigations for a potential deep geologic repository for spent nuclear fuel. The goal of the site investigation phase is to obtain a permit to build the deep repository for spent nuclear fuel. In parallel with the investigations, consultations will be held with county administrative boards, regulatory authorities and municipalities, as well as with members of the public.

The Nuclear Waste Management Organization of Japan (NUMO) has developed a set of “Siting Factors (SFs)” to guide choice of preliminary investigation areas (PIAs) of HLW disposal. A call for municipalities to volunteer PIAs was initiated, which included published SFs, in December 2002 as the first stage of a stepwise site selection procedure. This paper describes the way that SFs were developed and the outlines how they will be applied.

The selection of Preliminary Investigation Areas (PIAs) to be considered in the siting procedure for a Japanese High Level Radioactive Waste (HLW) repository, will require Site-specific Evaluation Factors (SSEF) to be considered. Evaluation of these factors requires a methodology for taking into account various kinds of uncertainties in varied types of literature data. The study described here evaluated the application of Evidential Support Logic (ESL) for this purpose. The approach involves constructing hierarchical process models. Uncertainties are then propagated from the lowest processes, corresponding to data or information, through intermediate processes, to some top level process of interest, such as “assessing the existence of an active fault”. To evaluate the usefulness of ESL a hypothetical site was assessed. The results demonstrate the value of the approach to support decision-making in the selection of PIAs.

The anticipated long term evolution of spent nuclear fuel as well as the remaining scientific key issues are presented for the various boundary conditions that can be encountered in long term dry storage and geological disposal. Spent fuel is expected to evolve significantly in closed system conditions which are representative of long term dry storage and the first stages of geological disposal. The mechanical evolution of the grain boundaries, the fate of helium and the evolution of the RN location within the pellet are the three major questions to be addressed which could significantly modify the physical and chemical state of the fuel. In addition, mechanisms and kinetics of fuel alteration by water in deep geological repository are still to be more deeply understood, in particular the inventory of the instant release and the radiolytic dissolution processes, to get a robust and reliable source term.

The Instant Release Fraction at container failure time, IRF(t), is here considered as being the sum of (i) the initial labile fraction, corresponding to the sum of gap and grain boundary inventories of certain radionuclides on exit from the reactor, with a further possible contribution from segregation in the rim region and (ii) the time-dependent fraction of radionuclides accumulating at grain boundaries due to a self-irradiation enhanced diffusion through the grains. The initial labile fraction of radionuclides such as 14C, 36Cl, 79Se, 129I, and 135Cs has been estimated based on leaching experiments, post-irradiation fission gas release measurements and studies of solid-state chemistry of spent fuel, along with estimates of fission product segregation in the rim zone. The contribution of the a self-irradiation enhanced diffusion has also been estimated based on a diffusion coefficient decreasing with time proportionally with the volume α-activity of the spent fuel. Its contribution to the IRF is limited for UO2 fuels. The proposed bounding values of the IRF for fuel with a burnup of 55 GWd/tIHM for 14C, 36Cl, 79Se, 129I, and 135Cs are 11 % at t=0 and close to 15 % at a container failure time of 10,000 y.

Interactions during extended exposure of UO2 to 2:1 H2+O2 mixtures at room temperature and 0.13 bar pressure are investigated in an effort to describe chemical and kinetic behavior of spent fuel following contact with groundwater. Oxidation of UO2 to UO2+x by O2 occurs initially when oxide is directly exposed to the gas mixture or submerged in water, but immersion is accompanied by a 25-fold reduction in the rate. The initial rate is proportional to [O2]2 for gasphase oxidation and to [O2]1.5 for the submerged oxidation. Continued measurement during direct oxide-gas contact indicates sequential reactions in which the UO2+x product is further oxidized by H2O and ultimately reacts with H2 to form an oxide hydride.

The objective of this work was to measure the actual rates of matrix dissolution of UO2 in synthetic groundwater by using an isotope dilution method to provide a quantitative estimate of precipitation effects. A preliminary series of tests was performed under oxidising conditions, in contact with air. The second series of tests was under reducing conditions produced by actively corroding iron. The solid phases were fragments of unirradiated fuel pellets and intact pellets. The aqueous phase was a dilute, synthetic groundwater - so-called modified Allard water that is buffered by sodium bicarbonate/carbonate. This paper gives results for the experiments under oxidising conditions and preliminary tests under reducing conditions. In contact with air, the U concentrations reached higher levels than measured in previous experiments with spent fuel or with UO2 pellets. The comparison of the U concentrations calculated from isotopicratios with the experimental results suggests precipitation has begun at later stages of restarted tests. The measurements in the presence of actively corroding iron gave very low concentrations in the aqueous phase. At contact times longer than one week, all U seemed to disappear from solution and sorb or precipitate on UO2 or Fe surfaces in the test vessel.

During long-term interim storage of spent fuel, pre-oxidation of the UO2-matrix may not be ruled out completely. This can happen if air could find access to the fuel in the case of cladding failure. The aim of this work is to study the impact of pre-oxidation of the fuel surface on the UO2 matrix dissolution rate and the associated mobilization or retention of radionuclides in highly concentrated salt solutions. The tests were performed with samples that suffered pre-oxidation during up to seven years. The dissolution rate of a fuel sample contacted by small quantities of air-oxygen was found to be roughly a factor of 10 higher in comparison to non oxidized samples, but concentrations of radionuclides, especially Pu and U were hardly affected. The majority of dissolved radionuclides, especially Pu, U appear to have been reimmobilized on the fuel sample itself.

To support the performance assessment (PA) calculations for the possible final disposal of spent fuel in the Boom Clay, leach experiments are performed with α-doped UO2 in clay media, simulating various near field ages. The experiments allow to measure the dissolution rate of the UO2 and to determine the assumed relationship between dissolution rate and α-activity. Tests are performed at six different α-activities, simulating various fuel ages, at 25–30°C, for durations ranging from 90 to 720 days, in a glove box with Ar/0.4%CO2 atmosphere. The solutions and solids are analyzed for U isotopes and 238Pu by radiochemical measurement and by ICP-MS. The dissolution rates of the α-doped UO2 are presented for different durations. The resulting corrosion rate is around 300 μgU.m−2.d−1. This is up to 100 times higher than found for similar conditions in the literature. In the presence of clay, there appears to be no correlation between the α-activity and the corrosion rate of the α-doped UO2.

A finite difference based modelling of the oxidation of uranium oxide is presented. This modelling involves only oxygen diffusion from the atmosphere into the solid and then inside the solid described with Fick's law. Some calculations performed with this modelling evidenced that the shape of the weight gain curves observed experimentally could be reproduced. This modelling is compared to the formula previously used to interpret the thermograms.

The estimation of the surface site density for UO2 (main component of the SF matrix) is an important aspect to take into account in the development of radiolytical models for the dissolution of the spent nuclear fuel (SF). Also, other oxides can be formed on the SF surface due to the effect of radiolytically-formed oxidizing species. Due to the lack or reliable data in literature we have studied the surface site densities of two uranium oxides: UO2 and U3O8.

For this determination, the knowledge of both the reactive surface area and the surface charge of the solid are necessary. In this work the reactive surface area of the two solids was determined by the BET method while the surface charge was determined by potentiometric acid-base titrations of suspensions of the solids at different ionic strengths (0.001, 0.01, and 0.1 mol·dm−3) under N2 atmosphere.

The amount of adsorbed protons was calculated by subtracting blank titration from thesolid suspension titration. The single-site nonelectrostatic model (NEM) was used to describe titration data. Uranium speciation in solution was included in the model as well.

The surface area values obtained were 0.15 ± 0.01 m2·g−1 for UO2 and 0.77 ± 0.02 m2·g−1 for U3O8, while the surface acidic site densities were determined to be 165 ± 10 sites·nm−2 and 48 ± 3 sites·nm−2 for UO2 and U3O8, respectively.

In Sweden, spent nuclear fuel will be encapsulated in sealed cylindrical canisters, consisting of a cast iron insert and a copper outer container. The canisters will be placed in a deep geologic repository and surrounded by bentonite. If a breach of the outer copper container were to occur the cast iron insert would undergo anaerobic corrosion, forming a magnetite film whose volume would be greater than that of the base metal. In principle there is a possibility that accumulation of iron corrosion product could cause expansion of the copper canister. Anaerobic corrosion rates are very slow, so in the work described in this paper reference was made to analogous archaeological materials that had been corroding for long periods in natural anoxic aqueous environments. This paper describes a number of archaeological artefacts containing iron and copper corroding in constrained geometries in anoxic natural waters. No evidence has been obtained which would suggest that severe damage is likely to occur to waste canisters as a result of expansive corrosion of cast iron under repository conditions.

In this work, the efficiency of one- and two-electron oxidants in oxidative dissolution of UO2 has been investigated. This was accomplished by measuring the U(VI)-concentration in solution after exposing UO2-powder to controlled amounts of oxidants in aqueous solution. The oxidants used in this study are H2O2, IrCl62−, CO3• and OH•. H2O2 acts as a two-electron oxidant while the remaining three oxidants are one-electron oxidants. CO3 and OH• were generated using γ-radiolysis. The results clearly show that the dissolution yields for one-electron oxidants (per electron pair) are lower than the yields for two-electron oxidants. Furthermore, the yields for one-electron oxidants increase with increasing amount of oxidant (especially at low amounts of oxidant). The rationale for this is that U(VI) is the main soluble species which is formed directly upon two-electron oxidation. For one-electron oxidants the primary oxidation product is U(V) which can form U(VI) upon reaction with a second oxidant. The probability for a second oxidation is however low at low concentrations of oxidant.

The dissolution of the unirradiated uranium dioxide, UO2, is largely affected by chemical variables like pe, pH and carbonate concentration in solution. In a previous paper [1], we studied the kinetics of the dissolution of UO2 as a function of pH and oxygen partial pressure and the experimental dissolution rates obtained showed a fractional dependence on both parameters. These fractional orders indicate a dissolution occurring via a surface mediated mechanism.

In this work we have developed a mechanism accounting for the oxidative dissolution of UO2, which depends on the oxygen concentration in solution and on the pH and a general rate equation has been deduced. This mechanism has been used to explain the results obtained by different authors and to explain and predict the dissolution of the spent nuclear fuel matrix.

The results of a spent fuel leaching experiment in which a fuel pin (17.7 g) was contacted with 380 mL of a 10 mM NaCl, 2 mM NaHCO3 solution by taking special care to minimize atmospheric oxygen contamination are presented. During the first 287 days, the fractions of inventory in the aqueous phase per day (f/d) increased nearly constantly for all nuclides (except for 100Mo), but were higher for fission products f/d(137Cs)=1.210−6, f/d(99Tc)=1.1·10−6 and f/d(90Sr)= 6.7 · 10−7 than for actinides: f/d (238U) =1.0 · 10−7, f/d(237Np)= 2.6 · 10−7 and f/d(239Pu) = 5.1 ·10−9. After adding iron, cast iron and copper foils (of ∼30 mm2 size), the concentrations of 238U, 237Np and 99Tc decreased by 80%, 97% and 88% to relatively stable levels (500ppb, 0.2 ppb and 0.6 ppb respectively). 239Pu concentrations increased from a level around 0.05 ppb to PuO2 solubility level, 0.5 ppb, and stabilized. The leaching process for 137Cs, 100Mo and 90Sr seems not to be influenced by the addition of metal foils. The observations in the present work contribute to an improved understanding of the behavior of spent fuel under near field repository conditions.

The dissolution rate of small UO2, ThO2 and mixed UO2–ThO2 particles, representative of unirradiated HTR fuel, are investigated in the pH range 1.7 to 5.5 under oxic conditions. For UO2 particles, the dissolution is kinetically control while for ThO2 particles, a solubility control mechanism seems to prevail for pH > 2. In the mixed oxides UO2–ThO2 particles, a selective release of uranium is observed, which could be either the result of a dissolution/precipitation mechanism in which thorium reprecipitate as oxide or hydroxide phase, or a result of a higher accessible surface area. Under oxic conditions, uranium may be considered as a tracer for the other soluble elements. So, assuming constant conditions with time, one can estimate from these preliminary results that unirradiated thorium based fuel kernels seem to present intrinsic confinement properties for 15000 years relative to soluble elements.

For the first time at ITU electrochemical quartz crystal micro-balance has been used for the investigation of UO2 thin film oxidation. The combination of cyclic voltammetry with nano-gram weight change measurements in-situ has shown at least two steps in the initial stages of UO2 dissolution under oxidising conditions relevant to spent fuel storage.

The interaction of Sr(II) with CSH phases has been investigated to distinguish adsorption from co-precipitation processes and to assess the relevance of these processes for performance assessments. Batch-type sorption experiments were carried out to study the adsorption of Sr(II) on synthetic CSH phases at different C:S ratios. Co-precipitation experiments were conducted by precipitating CSH phases in the presence of Sr(II) at two different precipitation rates.

Distribution ratios (Rd) of Sr(II) obtained from the sorption and co-precipitation studies were found to be similar in value. The results indicate that co-precipitation processes do not enhance Sr(II) uptake by CSH compared to adsorption. In both the co-precipitation and adsorption experiments the same sorption sites are accessible to Sr(II). The precipitation rate of the CSH phases has no significant influence on the Rd values.