Carbon steel is one of the candidate materials for overpacks for high-level radioactive waste disposal in Japan. Passivation behavior and corrosion rate of carbon steel were investigated by electrochemical measurements under simulated repository conditions. The results of the anodic polarization measurements showed that carbon steel was hard to passivate in highly compacted bentonite. Therefore, general corrosion seems to be most probable in repository conditions. In order to monitor the in-situ general corrosion rate in compacted bentonite, the AC impedance of carbon steel was measured under aerated conditions. It was confirmed that the corrosion rate in saturated bentonite decreased with time and it was almost the same as that observed in deaerated aqueous conditions. The corrosion rate did not increase in the presence of corrosion products formed by external current supply.

A series of batch laboratory experiments were set up to study the gross effects of microbial activity on repository geochemistry, radionuclide sorption and integrity of repository and host rock materials. The experiments were fully monitored for 550 days and some were continued, under apparently aerobic conditions, for 1700 days. Geochemical modelling of the experiments reproduced many of their features, but showed that the contents were not in chemical equilibrium after 1700 days with redox disequilibrium a likely feature. The models could not predict localised reactions such as enhanced pitting of steel which appeared to be microbially mediated.

After the USA decided in 1988 to no longer accept spent fuel elements from German material test reactors (MTR), a national back-end fuel cycle alternative was sought in the Federal Republic of Germany [1]. The aim is their direct final disposal in deep, stable geologic formations. The corrosion of material test reactor (MTR)-fuel element claddings (aluminium) in repository-relevant brines was examined. Before the aluminium cladding material can corrode, the POLLUX cask, containing the fuel elements, must be corroded. In this case, iron(II) and iron(III) ions are present in the brine. These ions decisively influence the corrosion of the MTR fuel element cladding material, therefore the mechanism responsible for this phenomenon should be identified. Tests were performed in which Fe(II) and Fe(III) salts were added to the brines. In these experiments, the percentage mass decrease of the aluminium cladding, the iron content of the brine, as well as the pH value were determined. As expected the results provided the information about the corrosion mechanism. The higher the concentration of iron ions in the brines, the higher the aluminium corrosion rate was for all three brines. Identical redox equilibria between Fe(II) and Fe(III) were formed in the brine, irrespective of whether Fe(II) or Fe(III) salt had been added. It is assumed that the acceleration of the corrosion rate is based on the fact that Fe(II) is reduced to metallic iron by absorbing the electrons produced during the oxidation of aluminium to Al(III). The aluminium cladding material does not function as a barrier for the release of radionuclides from the fuel elements. The results of this study show that the 0.38 mm thick aluminium cladding will corrode through after approximately four weeks.

Small angle neutron scattering from hydrating cement paste was compared to that from a paste made of a blend of cement and ground granulated blast furnace slag. The scattering measurements were used to determine the average size of the scattering particles during hydration. Their rate of growth was monitored in-situ over the first 40 hours of hydration. There was little difference in the scattering spectra, showing that the scattering particles in paste containing slag blend or cement were of similar size. At the beginning of hydration the particle radii were about 250 nn and after 14 days they had grown to over 600nm.

Estimates of particle sizes are critically dependent on the amount of multiple scattering. Therefore the amount of multiple scattering of the sample must be taken into account before making comparisons between the sizes of particles of different matrices.

Because of its experience in handling radioactive materials including waste generated on-site PSI was charged with managing the treatment and interim storage of the radioactive waste from medicine, industry and research in Switzerland. Combustible operational waste from the Swiss nuclear power plants is conditioned on a contract base and stored temporarily at the sites of the power plants. Because of the decommissioning of old facilities provoked by the change in the research priorities at PSI, temporarily an additional amount of waste is released on-site to be treated. PSI possesses the facilities and the organisational structure to fulfil the tasks it is charged with. In the near future the major part of the conditioning procedures, except the conditioning of some of the waste containing α-emitters, will be performed by ZWILAG. PSI will stay responsible for the management, organisation and interim storage of all radioactive waste, except for nuclear power plant waste, as before.

An innovative dry method for reprocessing highly enriched Al-U spent fuel is presented. It is based on the use of hot chlorine gas to chemically attack the irradiated fuels, converting aluminum, uranium and fission products into their respective chlorides. Separation of the large amounts of non-radioactive species, basically aluminum and its impurities, can be further accomplished by exploiting the significant differences between relative volatilities of metallic chlorides and also by selectivity of chlorination reactions. The technical feasibility of the method proposed is theoretically demonstrated by performing a case study on the highly enriched Al-U fuels of the RA-3 Argentine research and test reactor.

Our study is concerned with the different apatitic waste forms which can be produced with processes ranging from high-temperatures to ambient temperature synthesis. The different processes developped allow to take into account the nature of the waste and the application targeted, from specific conditioning of separated wastes, encapsulating of mixed wastes, solid or liquid wastes, to filling materials for direct storage of spent fuel, interim storage, surface storage or deep repository.

C-S-H of different Ca/Si ratios were synthesized in suspension. Cesium chloride (0.5 M) was put in contact with these C-S-H in reactors for 30 days at 25 °C with solution/solid ratios of 50. The quantities of cesium fixed by the C-S-H was determined by microanalyses and the mechanisms of retention in relation with the C-S-H structure was investigated by 133Cs Nuclear Magnetic Resonance (NMR). The influence of the humidity yield in which the C-S-H were stored was also studied.

At 100% of humidity, some cesium ions are trapped in the pore network of the C-S-H. However, most cesium ions are incorporated in the hydrated interlayers of the C-S-H between the silica chains of the structure. They are mobile in this interlayer space. At high Ca/Si ratio, the charge of incorporated Cs+ ions in the structure should be compensated by the associated Cl- incorporation. At low Ca/Si, the presence of silanol groups in the C-S-H structure, which becomes closer to that of tobermorite (Ca/Si about 0.8), makes possible the exchange or substitution H'<->Cs'. The resulting retention of cesium in the C-S-H then becomes higher.

A mixture of powdered magnesium-aluminum alloy (with composition close to Mg4A13), calcium silicide, and potassium nitrate as oxidizer was applied as metal fuel to incinerate various solid radioactive wastes (biomaterials, ion-exchange resin, etc.) and to produce glass-like materials. Materials obtained were examined with X-ray diffraction to identify mineral phases formed during incineration and vitrification processes. It has been established that solid waste incineration resulted in a formation of a slag product composed of stable mineral phases: oxyapatite, nagelschmidtite, spinels, karnegiite, which can be considered as host phases for radionuclide immobilization. Waste vitrification with metal fuel yields an amorphous phase which is predominant in final products, and a small amount of crystalline phases. Preliminary leaching tests were carried out. Leach rate of 137Cs in deionized water at 20 °C was found to be by order of magnitude 10−5 g-cm−2 day−1. Leach rates of 90Sr and 238Pu were approximately one order of magnitude lower.

As part of the evaluation of the safety of geologic disposal, the leaching behaviour of two types of bituminized radioactive waste (Eurobitum and CEA bitumen) has been studied as a function of temperature, pressure, leachant composition and bitumen matrix type. Inactive and active bitumen samples were brought into contact with two test media, simulating the geologic disposal environment of the Boom clay formation. At contact with these media, the samples swelled and soluble salts and radionucides were leached. It was found that the leach rate is influenced by temperature, the leachant composition, and the physical characteristics of the bitumen matrix. The release of nitrate is interpreted as a diffusion controlled process, which can however be disturbed by crack formation. The leaching of 60Co, 90Sr, and total β is diffusion controlled. Low leach rates were measured for Pu and Am: the release of Pu and Am is limited by their solubility in the leachant. Pu and Am are preferentially sorbed to the Boom clay, the test container or the bitumen. The leached mobile Pu and Am concentrations are of the order of 10−10 to 1013 M at 23°C. The results of this study suggest that the integrity of bituminized waste packages is seriously affected due to the leaching of soluble salts: a full-size 220 litre Eurobitum drum is predicted to be depleted in NaNO3 in less than 20,000 years.

The uptake of Eu(III) and Th(IV) by a sulphate-resisting Portland cement has been investigated in batch type sorption experiments using 152 Eu and 234Th as tracers. Prior to performing sorption studies, the experimental methodology was assessed. The kinetics of Eu(III) and Th(IV) interactions with cement were measured for times up to about 130 days. The effect of isosaccharinic acid added to artificial cement pore water on Eu(III) and Th(IV) uptake was examined to elucidate the effect of cellulose degradation products.

Flow and transport in a fractured rock is often conceptualized as flow between parallel plates and/or flow in tubes. For simple conceptual models, such as parallel plates or tube, it has been shown theoretically that the increase in effective porosity for transport is proportional to the variability (roughness) of the fracture aperture or tube radius. This paper focus on a tube model and estimates transport properties which are consistent with the estimated measured hydraulic properties from well tests in deep boreholes.

An alternative waste-package design for the geological disposal of high-level waste (HLW) glass is presented. In conventional designs, a massive buffer of compacted bentonite is placed around a thick-walled, mild-steel overpack; in the revised design, a much thinner buffer is placed within a thin-walled, mild-steel overpack. This simple expedient eliminates certain performance concerns in existing waste-package designs, while not necessitating the study of any new materials. This integrated waste package (IWP) design has comparable release-rate performance as current package designs for HLW. In addition, the 1WP design requires far-less rock excavation, permits significantly higher temperatures for longer periods, leads to a 20-50% reduction in repository area, and is more cost efficient than previous designs.

In a nuclear waste repository the disturbed rock zone adjacent to the deposition holes of waste canisters is a potential transport pathway of radionuclides. The properties of the disturbed zone may also play an important role in the saturation of the bentonite buffer and in gas release.

Three experimental holes of the size of deposition holes (depth 7.5 m and diameter 1.5 m) in a KBS-3 type repository were bored in hard granitic rock in the Research Tunnel at Olkiluoto. The disturbed rock zone caused by the boring was analyzed including determination of the porosity and structure of the disturbed and intact rock. These were studied in the laboratory using core samples by using the novel 14C-polymethylmethacrylate (14C-PMMA) method. In addition the structure of the disturbed zone was also examined using scanning electron and optical microscopy. A distinct zone of disturbance adjacent to the surface of the deposition holes was found. This disturbed zone was further divided into three separate sub-zones with gradual transition between the zones. It was also observed that the machine and tool factors have effect on the properties of the disturbed zone.

In order to be able to describe the migration of an alkaline plume, coupled chemistry and flow computer models may be employed. A series of laboratory column experiments were conducted to test the capabilities of currently available coupled models to predict product solids and output fluid compositions as a function of time. The predictions did not replicate all the variations of Ca:Si observed during mineralogical analysis, probably due the unavailability of sufficient kinetic and thermodynamic data for the range of calcium silicate hydrates (CSH) phases of interest. However, the predictions do reproduce the variation of Ca:Si ratios with time and distance along the columns. A better understanding of the above processes will lead to greater confidence in predictive modelling of the migration of alkaline plumes away from the cementitious barriers of radioactive waste disposal facilities.

The DRINK code is a 2D, biogeochemical transport code developed as a research tool to simulate the long term evolution of near surface LLW disposal sites and to generate gaseous and liquid source terms for far field studies. The code was recently upgraded to provide a more generic modelling tool with wider application to radionuclide migration scenarios. During the development of this code, the Generalised Repository Model (GRM), an integrated strategy has been employed to ensure the production of a fully tested, verified and quality assured product. This strategy is based around a code development protocol with three main components: quality assurance and documentation, verification and realism testing. Realism testing includes both peer review and model testing, with the latter including: experimental test cases; natural and anthropogenic analogues; field observations and finally uncertainty and sensitivity analysis. This paper describes the successful application of the protocol to the development and testing of the GRM code with specific emphasis upon verification and realism testing.

At the Finnish candidate sites for a nuclear waste repository calcite (CaCO3) is a common fracture mineral, that may participate in coprecipitation processes. The objective of this work was to study the coprecipitation of the trace element Ni with CaCO3 under controlled conditions. The experiments were carried out at 30 °C in vessels closed to the atmosphere. Calcite-saturated 0.05 M NaCI solutions containing trace amounts of Ni2+ were contacted with calcite for periods of up to 42 days. The experimental data indicate that Ni coprecipitates with calcite as a result of recrystallization. The amounts of coprecipitated Ni and recrystallized calcite were determined using liquid scintillation counting and the isotopes 63Ni and 45Ca. The results are supported by a complementary SEM/EDS analysis of the solid phase.

The chemical evolution of a clay engineered barrier in a deep repository interacting with various groundwaters is predicted using the TRIO-EF code supported by additional models tested against experimental data. Main results of this work are related to the long term evolution of the porewater and the buffering effect of the clay barrier in the presence of ion exchange sites and calcite, when exposed to representative granitic groundwaters or more acidic or alkaline waters.

A federal Environmental Assessment Panel has completed public hearings on the proposed concept for geological disposal of Canada's nuclear fuel waste. The Panel will make recommendations to assist the governments of Canada and Ontario in reaching decisions on the acceptability of the proposed concept and on the steps that must be taken to ensure the safe long-term management of nuclear fuel waste in Canada. It is instructive to review the background to the public hearings, to consider the issues that have been important in the public review, and to reflect on the opposing points of view presented at the hearings.

The application of a 3-dimensional geoscience information system (GSIS) in the performance assessment of the geological barrier of the nuclear waste disposal site Morsleben is presented. A primary objective is the development of a geometrically consistent three-dimensional model, representing all relevant hydrostratigraphic units in the model area. Following the completion of the structural model, Finite-Element mesh generation techniques are applied to derive a threedimensional spatial discretisation for numerical modelling of the regional groundwater movement.