[en] Common to all high-level-waste disposal concepts is the encapsulation of the waste into metal canisters. The purpose of this waste canister is to isolate the radioactive waste from contact with its surroundings for a desired time period. The design service life ranges from hundreds to thousands of years depending on the disposal concept. After the isolation has been breached, other barriers in the disposal system will delay and attenuate the radioactive releases to acceptable levels. In a deep geologic repository, the waste package will be exposed to chemical attack and, depending on the type of repository, to mechanical stresses. Each of these factors will by itself or in combination inevitably lead to loss of confinement some time in the future. In the design of the Swedish waste canister, the corrosion resistance is provided by an outer shell of pure copper while an insert supplies the mechanical strength cast nodular iron. The close fit between the insert and the copper results in very small tensile stresses in the copper over very limited areas once the repository has been saturated. Measurements of stress corrosion crack growth show that annealed copper cannot maintain sufficiently high stress intensity factors for cracks to grow. For annealed copper, the stress intensity factor was limited to 25 MPa·m^1/2 because of extensive plastic deformation. For cold-worked copper, no crack growth could be observed for stress intensity factors <30 MPa·m^1/2. Through the choices of canister material, canister, and repository design, and considering the expected chemical conditions, the risks for localized corrosion can be lowered to an acceptable level, if not eliminated altogether, and the releases from prematurely failed canisters can be kept well within acceptable dose levels

[en] Short fuel/clad segments from a high burnup PWR fuel rod have been exposed to simulated groundwater under both oxidizing and reducing conditions. Two methods of establishing reducing conditions were employed: by using H₂ gas in the presence of Pd catalyst, and by circulating the groundwater over rock-cores from a deep bore-hole. The results from the first two contact periods of 82 and 172 days are in good agreement with those obtained previously on a high burnup BWR fuel rod. In particular, it was found that under oxidizing conditions, uranium saturates at about the 1 mg/l level, and plutonium at about the 1 μg/l level. Under reducing conditions, these solubilities decreased by about two orders of magnitude. (orig.)

[en] Investigations have shown that fission products containing borosilicate glasses sometimes show a separated, yellow, alkali molybdate phase. It has been found that additives to the glasses as well as changes in melting conditions influence this phase separation. Knowledge of state and structure of molybdenum in glasses is, therefore, important. In this work simple silicate glasses containing MoO₃ were investigated by ESCA (electron spectroscopy for chemical analysis) using as a reference a simple phosphate glass. The Mo 3d spectrum in phase-separated and nonphase-separated glasses was investigated. The MoO₃ composition of the glasses varied from 2, 4, and 8%. Both oxidizing and reducing conditions were used for the 2% MoO₃ glasses. For the 4 and 8% glasses only oxidizing conditions were used. The recorded Mo-spectra are presented and results discussed

No abstract available

[en] Leaching of high burnup BWR fuel for up to 3 years showed that both U and Pu attain saturation rapidly at pH 8.1, giving values of 1-2 mg/l and 1 μg/1 respectively. The leaching rate for Sr-90 decreased from about 10^-5/d to 10^-7/d but was always higher than the rates for U, Pu, Cm, Ce, Eu and Ru. Congruent dissolution was only attained at pH values of about 4. The hypothesis that alpha radiolytic decomposition of water is a driving force for UO₂ corrosion even under reducing conditions has been examined in leaching tests on low burnup (low alpha dose-rate) fuel. No significant effect of alpha radiolysis under the experimental conditions was detected

[en] Solubility control for uranium is generally considered as a limiting factor for the release of soluble radionuclides from spent fuel. Under oxidizing conditions, the solubility-controlling uranium phase differs from the dissolving phase and a high thermodynamic driving force for UO₂ dissolution remains. Yet the experimental data show a decrease in the release rate of soluble radionuclides (i.e., ¹³⁷Cs, ⁹⁰Sr, ¹²⁵Sb) with time. In dynamic long-term tests, the initially high release rate of ⁹⁰Sr resembles the rate of uranium release with time. Since the uranium release is controlled by saturation in solution and by the frequency of solution exchange, the similarity of ⁹⁰Sr and uranium release means that with time uranium saturation controls the stability of the UO₂ matrix and the release of strontium. There seems to be the possibility of an agreement of the ⁹⁰Sr release data, structural information, and diffusion data for oxygen in UO₂. If this mechanism could be verified by comparison to other experimental data and by surface analytical techniques, the release of soluble radionuclides from spent fuel can be described as a transport process under moving boundary conditions. In the long run, a steady state should be achieved at which the flow rate and solubility-controlled release rate of uranium equals the diffusion-controlled transformation rate of the fuel matrix

[en] Spent fuel as a high level waste form has been studied for over a decade. Although the fuel contains a multitude of radionuclides, only a few of them will actually constitute a long term potential radiotoxic hazard. These elements are, after some thousand years, plutonium and neptunium. Up to about 100,000 years plutonium is by far dominating the radiotoxicity after which neptunium becomes the dominating element. In this paper, the results obtained on plutonium and neptunium releases from spent fuel are reviewed and discussed. Attempts are made to interpret the data obtained for different fuel types and groundwaters in terms of the chemistry of the respective elements as well as in terms of release kinetics

[en] The experiment configurations used in the Stripa in-situ test programme for HLW glasses are described. Results from in-situ testing of a HLW glass, similar to what is proposed for vitrification in UP2/3 plants at La Hague, are presented and compared with laboratory results for the same glass. The laboratory results for the glass - water system are found to be confirmed by the field data; after an initially high corrosion rate, the rate of attack drops considerably once silica saturation has been reached. This long-term rate was found to be about the same in the laboratory tests and in the in-situ tests. Also for the glass - bentonite - water system, the glass showed the same general behaviour in the in-situ tests as in the laboratory experiments, although the results obtained were less conclusive

[en] Current trends in modelling waste package performance are reviewed mainly from the perspective of the Swedish SKB studies. Examples are given, which illustrate the approaches for modelling different waste forms, i.e. HLW glass and spent nuclear fuel, and candidate canister materials, such as copper and steel. The relative importance of thermodynamics, reaction kinetics and near-field transport are discussed. 32 references, 7 figures, 1 table

[en] SKB has been evaluating a copper/steel canister for use in the disposal of spent nuclear reactor fuel. Once the canister is breached by corrosion, it is possible that the void volume inside the canister might fill with water. Water inside the canister would moderate the energy of the neutrons emitted by spontaneous fission in the fuel. It the space in the canister between and around the fuel pins is occupied by canister filling materials, the potential for criticality is avoided. The authors have developed a set of design requirements for canister filling material for the case where it is to be used alone, with no credit for burnup of the fuel or other measures, such as the use of neutron absorbers. Requirements were divided into three classes: essential requirements, desirable features, and undesirable features. The essential requirements are that the material fill at least 60% of the original void space, that the solubility of the filling material be less than 100 mg/l in pure water or expected repository waters at 50 C, and that the material not compact under its own weight by more than 10%. In this paper they review the reasons for these requirements, the desirable and undesirable features, and evaluate 11 candidate materials with respect to the design requirements and features. The candidate materials are glass beads, lead shot, copper spheres, sand, olivine, hematite, magnetite, crushed rock, bentonite, other clays, and concrete. Emphasis is placed on the determination of whether further work is needed to eliminate uncertainties in the evaluation of the ability of a particular filling material to be successfully used under actual conditions, and on the ability to predict the long-term performance of the material under the repository conditions