The effects of the temperature and high radioactivity on the chemical and physical state of the fuel during a confinement phase were evaluated in the framework of the PRECCI program. No significant evolution of the fuel chemistry is expected. The radionuclides mobility due to thermal diffusion can be neglected in the temperature range of disposal and diffusion enhanced by α self-irradiation appears to be limited. Microscopic swelling due to alpha self irradiation cumulated damage corresponds to a low volume increase of about 1%. Macroscopic swelling will depend on the helium behavior in the spent fuel. Literature data and results from the project indicate that helium solubility is low in UO2 and should be reached after a few centuries of cooling for MOX fuels. Although the thermal diffusion of helium is some orders of magnitude faster than the thermal diffusion of xenon, it remains low in the temperature range of disposal. Therefore, the first step of physical damage due to helium production should be the precipitation of He bubbles in grains, causing micro-cracks, which could enhance the He release to the grain boundaries. These processes and their consequences on the stability of grain boundaries still need to be quantified today.

The EIG EURIDICE is responsible for performing large-scale tests, technical demonstrations and experiments so as to assess the feasibility of a final disposal of vitrified radioactive waste in deep clay layers. This programme is part of the Belgian Research and Development programme managed by ONDRAF/NIRAS. The research infrastructure includes the Underground Research Facilities HADES (URF HADES) in the Boom Clay geological formation and surface facilities. The achievements of the demonstration programme are the demonstration of the construction of shafts and galleries at industrial scale, the characterisation of the hydro-mechanical response of the host rock, and the “OPHELIE mock-up” a large scale hydration test under thermal load of pre-fabricated bentonite blocks. The future works will consist mainly in the realisation of the “PRACLAY experiments” including a large scale heater test. The large scale heater test has to demonstrate that Boom Clay is suitable, in terms of performance of the disposal system, to undergo the thermal load induced by the vitrified waste. The combined effect of the excavation and the thermal load will be investigated. A long term (more than 10 years) large scale heater test would be representative of the most penalizing conditions that could be encountered in the real disposal. The results of this test will constitute an important input for the Safety and Feasibility Cases 1 (SFC-1, 2013) and 2 (SFC-2, 2020).

Steel components in the engineered barrier of HLW repositories have a potential to react with the clay buffer and induce adverse effects on its properties. These effects have been assessed within the KBS-3H disposal concept, which includes spent fuel in copper canisters surrounded by bentonite blocks and a steel supercontainer (SC), placed within tunnels in deep crystalline Olkiluoto-type host rock. Based on literature review and thermodynamic modeling, the SC is predicted to completely corrode within a minimum of a few thousand years. The resulting corrosion products are expected to be magnetite, and depending on groundwater compositions, also iron sulfide and perhaps siderite. Furthermore corrosion-derived Fe(II) may react with the clay to form non-swelling phyllosilicates, such as berthierine. From simple mass balance estimates and by applying a more realistic reactive transport model, it can be deduced, that the effect of iron from the SC on the integrity of the bentonite buffer is limited. However, large uncertainties in adequately quantifying this effect pertain due to limited process knowledge on Fe-bentonite interactions.

Since reprocessing is no longer the reference policy in Belgium, studies on the direct disposal of spent fuel in a clay formation have gained increased interest in the last years. In order to determine to what extent the clay properties and the α-activity influence the dissolution kinetics of spent fuel for the long term disposal, static dissolution tests have been performed on 5 different types of α-doped UO2, representing a PWR fuel with a burn-up of 45 or 55 GWd · tHM−1 and fuel ages ranging between 150 and 90,000 years, in different Boom Clay (BC) media at room temperature and in an anoxic atmosphere for 90 to 720 days. The uranium activity in the clay fraction over time was found to be much higher than the U-activity in the leachates, which has been mainly ascribed to the high retention capacity of the BC. The average dissolution rate between 0 and 90 days obtained for the 5 types of α-doped UO2 were all found to be high and quite similar at ∼263 µg · m−2· d−1and a “longer-term” rate (181 to 720 days) ranging between zero and 15 µg · m−2· d−1. These results suggest that the activity of the fuels does not seem to have an effect on the UO2 dissolution rates under the considered test conditions. In order to study the partition/redistribution of U during UO2dissolution, sequential extraction experiments were performed. Results of the latter have provided a better mechanistic understanding of BC/spent fuel interaction processes as well as information on the role of the different minerals controlling the U-retention/immobilization.

We discuss quantitatively the capability of direct α-recoil to create observable radioactive disequilibrium (U-234/U-238 activity ratio > 1) in flowing groundwater. Coupled equations for radioactive decay chain are formulated for the solid U source and for the corresponding groundwater receiving the recoiling U-234 atoms. U-rich fracture coating material and the water flowing in the fracture are discussed in detail.

The novel modelling approach worked well and provides a useful tool for more detailed site-specific studies with more detailed input data. The approach appears feasible because the equations are based on straightforward mass balance considerations.

α-recoil -induced enrichment of U-234 in groundwater is a slow process: the increase of U-234/U-238 activity ratio to notable values above unity will take hundreds to thousands of years. Strong groundwater flow prevents any local recoil-induced U-234 enrichment in the water.

In the horizontal emplacement concept (KBS-3H) for the disposal of radioactive waste, which is being developed in Sweden and Finland, copper canisters with cast iron inserts will be surrounded by bentonite buffer and mounted in perforated carbon steel support structures in boreholes within the bedrock. The groundwater will be reducing, leading to anaerobic corrosion of the ferrous material. It is important to understand both the effect of bentonite on the corrosion behaviour of the steel and the effect of the corrosion products on the performance of the bentonite. Previous work on the corrosion of steel in bentonite was extended to investigate a wider range of conditions, including the possible effects of alkaline plumes released from concrete support structures and the effect of chloride concentration and temperature on the corrosion rate of steel in bentonite. Corrosion rates were measured by collecting hydrogen produced by the anaerobic corrosion of iron. In addition, a range of analytical techniques was applied to study the composition and morphology of the corrosion products and the distribution and chemical state of the iron released into the bentonite. Comparison was also made between corrosion in compacted bentonite and artificial bentonite porewater. In the presence of bentonite, the corrosion product layer was relatively thin compared to fully aqueous conditions, probably because the ferrous ions released by corrosion exchanged with the bentonite interlayer or attached to the surface of the bentonite grains, rather than forming a separate iron oxide phase.

The apparent diffusivities (Da) and activation enthalpies (ΔEa) for I− and Cs+ ions in compacted Na-smectite were obtained in the parallel and perpendicular directions to the orientation of smectite particles as a function of smectite's dry density (0.9-1.4 Mg/m3), salinity ([NaCl]=0.01, 0.51 M) and temperature (295-333 K). The Da-values for both ions tended to be higher in the parallel direction than in the perpendicular direction to the orientation of smectite particles. The Da-values of I− ions in the parallel direction decreased with increasing salinity only at low dry density, but those of Cs+ ions increased with increasing salinity for all conditions. Based on this, it is interpreted that I− ions predominantly diffuse in external pores and Cs+ ions diffuse in both interlayer and external pores. The ΔEa-values forI− ions are at similar levels as those for the ionic diffusivity in free water (D°) of I− ions at low dry density and increased with increasing dry density. The ΔEa-values for Cs+ ions are higher than those for the D° of Cs+ ions even at low dry density and increased with increasing dry density. Such high ΔEa-values for Cs+ ions are considered to be due to the combined effects of the ion exchange enthalpy between Cs+ and Na+ ions in smectite and the lowering in the activity of porewater.

Burnable radioactive waste management options are often in favour of the vitrification due to their volume reduction, resulting in cost saving for transport and storage. More, the interest is also to obtain a best radioactivity immobilization in a long-lasting matrix. Development made on incineration-vitrification process must take into account radioactivity environment in term of process compactness, gas treatment and secondary wastes. A new and simplified equipment, using thermal oxygen plasma coupled with a cold crucible melter, is being developed at the Valrho-Marcoule nuclear centre (France) : SHIVA, for Advanced Hybrid System for Incineration and Vitrification. As part of studies concerning the management of bitumen canisters stored in Marcoule, the feasibility of their treatment in an incineration-vitrification process have been tested. This treatment could concern the most radioactive bituminised waste drums because they have to be sent in a costly storage site in consideration of their radioactivity. The interest of such treatment is first to reduce the volume of wastes in order to appreciably diminish the cost of this storage.

The safe disposal of radioactive nuclear wastes using multi-barrier systems requires a good understanding of their transport properties in geological materials. When groundwater flow is very slow, the most important mechanisms of transport and retardation are diffusion through and sorption onto geological materials, and an evaluation of relevant parameters, particularly the effective diffusion coefficient and rock capacity factor (or sorption coefficient), is of fundamental importance to any safety assessment. Although laboratory diffusion tests can be used to determine these two parameters, conventional through-diffusion testing has some limitations, such as the need for a relatively long test time, cumbersome test procedures and the possibility of errors due to differences between analytical assumptions and actual test conditions. In this paper, we offer a rigorous solution to the through-diffusion test. Boundary conditions are improved by taking into account concentration changes in both the source and the measurement cells. A companion approach for back-calculating the effective diffusion coefficient and rock capacity factor is also proposed. The effectiveness and advantages of this improved technique are demonstrated using experimental data derived from a sedimentary rock sample taken from a research site in Japan.

Vitrification is a potential route for the immobilisation of Plutonium Contaminated Material (PCM). This is an Intermediate Level Waste (ILW) arising from operations in which there is contact with Pu isotopes. PCM consists of low levels of Pu combined with metals, masonry, glass, ceramics, polymers and other carbonaceous materials. Simulated PCM containing CeO2 as a PuO2 surrogate was mixed with a phosphate precursor and vitrified. Pre-oxidation of PCM simulant prior to vitrification minimised the violence of batch reactions. No pre-oxidation produced inhomogeneous slag-like materials with high residual metals and particulates. Pre-oxidation at 600°C in air and at 1200°C in an O2-rich atmosphere produced more favourable results, with increasingly vitreous products resulting from more oxidised PCM simulant. The most oxidised PCM simulant produced phosphate glasses with low levels of particulate inclusions, as confirmed by x-ray diffraction and scanning electron microscopy. Particulates included iron-rich metallics and aluminous oxides. Increased melting times and temperatures may have reduced the number of inclusions slightly, but O2 bubbling during melting resulted in little additional benefit. Waste loading equivalent to ∼60 weight % of untreated waste may be possible. There was little evidence of Ce partitioning, indicating that it was immobilised within the glass matrix and had little preference for metallic or crystalline phases. These results demonstrate the potential feasibility for vitrification of PCM in phosphate glass, justifying further investigation into this potentially novel solution.

Concrete materials in nuclear facilities may become activated or contaminated by various radionuclides through different mechanisms. Consequently, decommissioning and dismantling of these facilities produce considerable quantities of these materials (e.g. concrete structures, rubble), which are at least potentially contaminated with radionuclides and which must be managed safely and cost-effectively. In this paper, we present results from a research project that aims at the development of source-term models for the mobilization of radionuclides from contaminated concrete. The objective of this task was to clarify whether a more realistic source-term description could be beneficial for optimization of the management of decommissioning wastes by reducing the amount of material for disposal as radioactive waste as well as by saving natural resources due to the recycling of building materials.

To identify important parameters and processes that affect the release rates of radionuclides, we evaluated the chemical behavior and the solid speciation of radionuclides in concrete materials and the influence of factors like concrete properties, source/pathway of contamination, and the scenario-specific chemical environment and hydraulic regime. Furthermore, concrete degradation processes and their influence on contaminant mobilization were addressed. On this basis, source-term models were developed to describe the radionuclide release by (i) the dissolution of radionuclide containing solid phases, (ii) the desorption of radionuclides from surfaces, and/or (iii) the leaching of radionuclides from a solid matrix without disrupting its structure. These source-term models were parameterized for probabilistic simulations of various release options, including the reuse of recycled building materials, the disposal of rubble in inert and municipal landfills as well as the on-site disposal of concrete materials (e.g. foundations remaining in the ground, in situ burial of rubble). For some scenarios and radionuclides, the calculated release rates were between one and two orders of magnitude lower than those used in former generic calculations. Based on the results of stochastic simulations, the consequences of the use of a more realistic source-term for dose assessments with respect to clearance/recycling of contaminated concrete will be illustrated and discussed.

The corrosion resistance of container materials in underground repositories is an important issue for the safe disposal of High Level Nuclear Waste (HLNW). The reliable prediction of container degradation rate and engineering barrier integrity over extended periods, up to several thousands years or even several hundreds of thousands of years, represents one of the greatest scientific and technical challenges. The first and the second International Workshops on Prediction of Long Term Corrosion Behaviour in Nuclear Waste Systems, which were held in 2001 (Cadarache) and 2004 (Nice), sought to compare the scientific and experimental approaches that are being developed in various organisations worldwide for predicting long term corrosion phenomena, including corrosion strategies for interim storage and geological disposal. The lessons learned during these Workshops, include the necessity of developing two approaches based on semi-empiricism and determinism in a complementary manner for effective prediction. The use of archaeological artefacts to demonstrate the feasibility of long term storage and to provide a database for testing and validating modelling work was also emphasized.

The amounts of cesium and uranium released from crushed spent PWR fuel in the gel-state clays with a few ml of supernatant at hot cell temperature under Ar-atmosphere have been measured. The fractions of cesium dissolved from the fuel for 873 days were 0.29 and 0.25% in Boom clay/Boom-clay water and Ca-bentonite/synthetic granitic groundwater, respectively. These cesium fractions were very close to the gap inventory of cesium, which was determined to be around 0.30% in the previous experiment. The fraction of uranium released up to 193 days in the Boom clay media was 0.011% and this fraction has been retained until 873 days. Such this phenomenon was also obtained in the Ca-bentonite media even though the released fraction was higher than that in Boom clay. The increase of less than 0.001% in the dissolved uranium fraction between 193 and 873 days suggests that the long-term leach rate of uranium from spent fuel would be much less than 24 μg·m−2·day−1.

The ion exchange phase of corrosion of nuclear waste glasses was modelled using Doremus' theory of interdiffusion and numerically analysed for British Magnox waste and Russian K-26 glasses. It is shown that even in non-silica -saturated conditions the ion exchange phase plays a significant role in the overall radionuclide release inventory particularly for short-lived radionuclides.

The iron phosphate cement (IPC) structure formation process has been studied by Mossbauer spectroscopy. The IPC is formed as a result of interaction of iron oxides with orthophosphoric acid and could be used as a matrix for immobilization of low-level radioactive waste.

The structural formation process has been shown may be considered to consist of two stages. The first stage lasts to the moment of setting, when the diffusion process goes on in the liquid phase. At this stage the main fragments of the structural polymeric frame of the IPC are developed consisting of iron (+2) and (+3) phosphates. The rate of the oxides-with-phosphoric-acid reaction as well as the time of hardening depends on the bivalent iron content.

The second stage begins from the moment of setting when the diffusion becomes slower. At this stage the process is characterized by the negligible increase in the iron phosphates (+2), (+3) content and transformation of the previously formed phosphates.

The nuclear gamma-resonance (NGR) parameters have been determined of FeH3(PO4)2·2.5H2O, forming in the hematite (Fe2O3) based IPC: isomeric shift (IS) = 0.46 mm/s, quadrupole splitting (QS) = 0.197 mm/s, FWHM =0.282 mm/s.

Technetium was immobilized in a Mg2TiO4 matrix by atomic substitution of titanium for 20, 50 and 80mol%. Leach tests, conforming to the ASTM C-1220-92 specifications, were carried out using simulated leachants (granite water, clay pore water and saturated salt solution) in order to determine normalized leach rates (NL) and fractional releases (FR%).

FR% for technetium have been measured < 0.1% for all Tc-loadings. Typical NL are1.7 E−2 g/m2d and 1.1 E−1 g/m2d for a 50 mol% and 80 mol% Tc-loading. In comparison to published fractional releases for glass, the matrix proves to be a more suitable host.

In addition, inactive pellets were produced and leached for investigation of the behaviour of the matrix material itself and the influence of other phases due to possible incomplete reactions during fabrication. Two types of inactive pellets were produced: Mg2TiO4 and Mg2TiO4/MgTiO3/MgO mixed pellets, the latter by decreasing the temperature during fabrication to 1285°C. Release of titanium from MgTiO3 proves to be higher than for Mg2TiO4.

It was found that the leachability of the host matrix increases upon increasing the technetium loading. As a compromise between waste volume reduction and leachability, it is suggested that technetium loadings do not exceed about 50 mol%.

Chemical durability of spent fuel under repository conditions is one of the main topics of interest in national and international projects from the last two decades. During the last decade there have been growing the activities in Enresa related to the deep disposal concept with the aim of developing a spent fuel alteration model for the understanding of the behaviour of this nuclear waste under repository conditions. In this context, the development and utilisation of models and sub-models based on experimental work have been of great importance.

Experimental studies with spent fuel(in Collaboration Agreement with ITU) and spent fuel analogues in several environmental conditions have been carried out into Enresa R+D Programmes in order to reach a better knowledge of the relevant processes and to quantify the spent fuel chemical durability under repository conditions.

In this work, it is showed the utility of data provided from experiments with spent fuel analogues to test the mechanisms and the influence of relevant parameters in the spent fuel alteration under repository conditions.

The evolution of irradiated fuel under interim storage conditions and in deep geologic storage and its oxygen to metal ratio (O/M) before the water access to the fuel is another factor of great influence on enhanced spent fuel leaching. This effect has been also studied by means of spent fuel analogues and by using simulated (artificially) aged fuel.

Experiments were performed in anoxic gloves box in an attempt to synthesise Coffinite both in representative near-field conditions, and in conditions which were expected to favour its precipitation according to thermodynamic calculations. The experimental results did not confirm the predictions. However, a new mineral was observed instead of Coffinite. In addition, accurate characterization of various natural samples demonstrate the permanent presence of U(VI) within Coffinite contradictory to its theoretical composition. Our observations raise the question on the validity and applicability of available –actually estimated- thermodynamic data of Coffinite. Based on kinetic hindrance of Coffinite formation, coffinitization of spent nuclear fuel in geological disposal is not anticipated to be a dominant short term process.

In accordance with the R&D framework specified by the Atomic Energy Commission of Japan in 2000 for the implementing phase of HLW disposal, the Japan Nuclear Cycle Development Institute (JNC) continues to be responsible for R&D activities aimed at enhancing the reliability of disposal technologies and safety assessment methodologies and associated databases. JNC has thus been actively promoting technical R&D with a view to contributing to both the implementation of disposal and the formulation of safety regulations.

One of JNC's key roles is to establish and demonstrate site characterization methodologies based on investigations in two purpose-built generic URL (underground research laboratory) projects: one at Mizunami in crystalline rock and the other at Horonobe in sedimentary rock.

Through the surface-based investigations in the Mizunami and Horonobe projects (phase 1), integration of work from different disciplines into a “geosynthesis” has been illustrated and is planned to be developed further in the underground facilities at these sites (phases 2 and 3). These projects also serve for developing and testing the tools and methodologies required for site characterization. Further know-how will be gained through participation in foreign underground laboratory projects, transfer of experience from these projects to Japan and tailoring it to Japanese conditions and requirements.

This experience represents an important knowledge base, which is obviously important for the implementer but is also needed by the regulator, in order to assess how key site characteristics are derived and what uncertainties are associated with this process.

To ensure safety and efficiency of construction and operation of repositories, rock fractures need to be grouted where they intersect the facility. Cement-based grouting materials that have been used extensively in the past may, however, react with the groundwater and produce hyperalkaline leachate. A possible solution for this issue is use of alternative material.

Ethanol/Bentonite Slurry Grouting is another approach to minimize the long-term permeability of rock fractures. The slurry has high fluidity for the case of the ethanol concentration in the liquid phase being 60%. The viscosity of the slurry with bentonite content of 0.4 Mg/m3 was measured for conditions of the shear rate ranging from 0.1 to 1000 1/s. The viscosity of the slurry changed dramatically between the ethanol concentrations of 60 to 40 %, behaving as a non-Newtonian fluid. In order to investigate when the ethanol content of the slurry becomes low enough to achieve sufficient viscosity in rock fractures, an Ethanol Diffusion Test with experimental openings was carried out in the laboratory. Extrapolation of the measured minimum time for the filled slurry to reach the desirable low concentration of ethanol lead to the estimated minimum time ranging from 0.1 hours to 0.7 hours for a 0.2 mm aperture. These facts suggest that the injected slurry flows through the openings of fractures during the early period of injection and, later, as its ethanol content decreases, slurry's viscosity increases sufficiently so that the openings can be blocked.