To experimentally assess the long term creep performance of oxygen-free phosphorus- doped (OFP) copper for the overpack of repository canisters, the combination of modestly elevated temperature and multi-axial stress state has been applied for accelerated testing. Multi-axiality was induced by using notched compact tension (CT) specimens, with interrupted testing to periodically inspect for creep damage. Uniaxial creep testing was also conducted to support creep analysis of the CT specimens. After about 10000 h of testing at 150°C/46 MPa (reference stress), the inspected CT specimens showed only marginal creep cavity indications near the notch tip. However, a distinct grain boundary zone with elevated Pcontent was observed to appear and widen during testing, mainly near the notch tip. The significance of the grain boundary zone is not well understood, but indicates stress-enhanced microstructural changes at relatively low temperatures. The predicted isothermal uniaxial creep life at 150°C/46 MPa agreed satisfactorily within a factor of two in time, when obtained independently from converted multi-axial testing results and directly from a creep model based on the available uniaxial data. Although the uncertainties in extended extrapolation remain large, the prediction would suggest safe long term service at leastagainst pure creep failure of intact parent material.

The long-term immobilization for individual radioelements released from the waste form “spent fuel” in solid phases upon groundwater contact depends strongly on the (geo)chemical constraints prevailing in the repository. Related experimental studies comprise effects induced by the presence of Fe based container material, and near field materials other than Fe for a rock salt environment. The effect of the presence of an argillaceous host rock containing organic matter and pyrite on fuel alteration was studied in addition. The results have shown that oxidative radio-lysis products were found to be consumed at a significant extent by the metallic Fe and by the argillaceous host rock. Under these conditions a decrease at a factor of ca.100 for both the matrix dissolution rates and the solution concentrations of U and Pu was found. There is mutual support between the matrix dissolution rates, the solution concentrations and the amounts of oxygen encountered during the experiments under various conditions controlled by the presence of near field materials under study.

Crystalline rocks have fractures which may cause unexpected routes of groundwater seepage. Cement grouting is one of the most effective methods to minimize seepage; however, cement materials may not be suitable for the purpose of extra-long durability, because cement is neutralized or degraded by chemical and physical influence of chemical reaction.

Natural clay like bentonite is one of the most promising materials for seepage barrier; however, water/bentonite grout is so viscous that enough amount of bentonite can not be grouted into rock fractures. To increase bentonite content in grout with low viscosity, the utilization of ethanol as a mixing liquid was studied. Ethanol suppresses bentonite swelling, and more bentonite can be injected more than that of water/bentonite slurry. In this paper, grouting into in-situ rock mass fracture from the ground surface was tested to investigate the barrier performance and workability of ethanol/bentonite slurry as a grouting material.

The objective of this study is to clarify the mechanical effect on the engineered barrier system (EBS) of a fault movement, presupposed to occur in a high-level radioactive waste repository. The plan of this study is; 1) to understand the mechanical behavior of the buffer material when shearing takes place during experimental tests, and 2) to make progress in numerical analysis techniques in order to estimate the effect of fault movement on the EBS at a disposal site.

Accordingly, the first part of this paper reports tests, which were carried out on a 1:20 model of the EBS. The experimental results indicate that the increased total pressure is due, in part, to the increase of pore water pressure. The second part of this paper reports the results of finite element simulations of the experiments. The calculation results show that the permeability of the near-field rock, which influences the amount of water draining from the buffer material, affects the pressure increase in the buffer material. With appropriately set parameters, the calculation shows agreement with the experimental results.

Titanate-based ceramics with yttrium, gadolinium, manganese, uranium, calcium, aluminum, and iron oxides being of interest as matrices for radioactive wastes containing lanthanides, actinides and corrosion products were synthesized using cold pressing and sintering and melting routes. Major constituents of the ceramics are fluorite structure phases (pyrochlore, “murataite”, cubic oxide with fluorite-related lattice) as well as brannerite. Incorporation of aluminum and iron oxides (corrosion products) forms second phases with perovskite or/and crichtonite structure.

ONDRAF-NIRAS has recently selected a Supercontainer with a Portland Cement (PC) buffer as the preferred new reference design for disposal of HLW and spent fuel. The selection process involved a multi-criteria analysis of alternative design options, which were evaluated against a range of long-term safety and feasibility criteria. A PC concrete has been chosen for the buffer because this will provide a highly alkaline chemical environment, which will last for thousands of years. In this environment the external surface of the overpack will be passivated and overpack corrosion will be inhibited. The concrete buffer also has low-hydraulic conductivity to slow the infiltration of external fluids to the overpack surface, and provides radiological shielding.

ONDRAF-NIRAS has made a preliminary evaluation of the viability of the reference Supercontainer design. The following areas were reviewed and investigated: radiolysis, thermo-hydraulic (TH) behaviour of the concrete buffer, metal corrosion, the chemical and mineralogical evolution of the concrete buffer, and relevant industrial experience. This paper describes the main findings, and identifies remaining design and performance uncertainties. Prioritisation and recommendations for future work are also given.

This paper reviews the main conclusions of two European Commission funded projects, GLASTAB and GLAMOR, on the durability of HLW glass in geological disposal, and the remaining uncertainties. The progress in GLASTAB relates to the characterization of the alteration layer, the modeling of glass dissolution, the interaction of glass with near field materials, the behaviour of radionuclides, and the calculation of HLW glass dissolution as part of geological disposal. The GLAMOR programme was focused on the decrease of the glass dissolution rate upon achieving silica saturation in solution. Two conclusions from this programme are the importance of the residual rate measured beyond silica saturation and the need to reduce the uncertainties in model parameters. The overall conclusion from the two projects is that strong progress has been achieved on basic dissolution mechanisms and interaction with disposal environments - HLW glass will act as a good barrier to the release of radionuclides in geologic disposal.

A two-dimensional numerical model incorporating a multiple-canister configuration and a non-uniform flow field of the host rock is developed for more reliable performance assessment of a water-saturated high-level radioactive waste repository. A space-dependent flow velocity distribution is established numerically using the finite-element method. Transport of Cs-135 is simulated using the random-walk method. Numerical results for a source condition of instantaneous-pulse-input show that (i) migration of Cs-135 is influenced significantly not only by the canister configurations but also by flow boundary conditions, ii) effect of applied hydraulic gradient direction on Cs-135 migration becomes more significant as the number of canisters increases, while the effect is negligible for a single-canister configuration, (iii) the direction of the hydraulic gradient orthogonal to the orientation of the disposal tunnel is most advantageous with respect to the isolation of the radionuclide, and (iv) the single-canister configuration yields conservative results compared with multiple-canister configurations.

A combination of electrochemical, scanning electrochemical microscopy (SECM), X-ray photoelectron spectroscopic (XPS), and time of flight secondary ion mass spectrometry (TOF - SIMS) techniques are being employed to investigate the effects of dissolved hydrogen on the aqueous corrosion of uranium dioxide (UO2) under nuclear waste disposal conditions. Corrosion potential (ECORR) measurements indicate that the oxidation of dissolved hydrogen on noble metal ε-particles polarizes the UO2 nuclear fuel surface to reducing potentials (−300 – −400 mV vs. SCE; i.e., to ECORR values more negative than those observed under anoxic (argon-purged) conditions (−200 mV vs. SCE). A comparison of the behaviors observed on SIMFUEL specimens with and without incorporated noble metal ε-particles indicate that these particles may act as catalytic electrodes for H2 oxidation. It is the galvanic coupling of these particles to the UO2 matrix which suppresses the fuel corrosion potential thereby preventing oxidation of the fuel surface.

The strategic approach and structuring of safety cases for nuclear waste repositories have seen marked improvements over the past ten to fifteen years as a result of the collective experience of national agencies in developing their safety cases and their efforts, under the auspices of the OECD Nuclear Energy Agency (NEA), to develop a systematic and thorough approach, as well as through systematic reviews by regulators and expert teams. A safety case can be developed at any stage in a program and it provides a remarkably effective method for reassessing design and siting requirements and refocusing R&D needs. The process of developing a safety case for a repository is the critical test of the quality of understanding that exists for a given repository concept, because it forces an integration of the siting and design strategies and the assessment approach. This integration demands that one evaluates all relevant R&D results available and addresses the questions i) is the system good? and ii) is the understanding sufficient, taking the stage of the programme into account?

Here we address some questions on the relationship between R&D work and the development of a safety case, including:

1) Is the R&D performed in Europe evolving in concert with safety case development for European repositories?

2) How much generic work is still needed when robust site-specific repository concepts obviously lead to much more focused R&D?

3) How good is good enough and how do we measure this at each successive stage of repository development?

There are no easy answers to such questions, but they must increasingly be addressed as national programs progress.

The long-term behaviour of vitrified high-level waste in an underground clay repository was assessed by using the reactive transport model HYTEC with respect to silica diffusion, sorption and precipitation processes. Special attention was given to the chemical interactions between glass, corroded steel and the host-rock considering realistic time scale and repository design. A kinetic and congruent dissolution law of R7T7 nuclear glass was used assuming a first-order dissolution rate, which is chemistry dependent, as well as a long-term residual rate. Without silica sorption and precipitation, glass dissolution is diffusion-driven and the fraction of altered glass after 100,000 years ranges from 5% to 50% depending on the fracturation degree of the glass block. Corrosion products may limit glass dissolution by controlling silica diffusion, whereas silica sorption on such products has almost no effect on glass durability. Within the clayey host-rock, precipitation of silicate minerals such as chalcedony may affect glass durability much more significantly than sorption. In that case, however, a concomitant porosity drop is predicted that could progressively reduce silica diffusion and subsequent glass alteration.

U6+−phases are common alteration products of spent nuclear fuel under oxidizing conditions, and they may potentially incorporate actinides, such as long-lived 239Pu and 237Np, delaying their transport to the biosphere. In order to evaluate the ballistic effects of α-decay events on the stability of the U6+−phases, we report, for the first time, the results of ion beam irradiations (1.0 MeV Kr2+) for six different structures of U6+-phases: uranophane, kasolite, boltwoodite, saleeite, carnotite, and liebigite. The target uranyl-minerals were characterized by powder X-ray diffraction and identification confirmed by SAED (selected area electron diffraction) in TEM (transmission electron microscopy). The TEM observation revealed no initial contamination of uraninite in these U6+ phases. All of the samples were irradiated with in situ TEM observation using 1.0 MeV Kr2+ in the IVEM (intermediate-voltage electron microscope) at the IVEM-Tandem Facility of Argonne National Laboratory. The ion flux was 6.3 × 1011 ions/cm2/sec. The specimen temperatures during irradiation were 298 and 673 K, respectively. The Kr2+-irradiation decomposed the U6+-phases to nanocrystals of UO2 at doses as low as 0.006 dpa. The cumulative doses for the pure U6+-phases, e.g., uranophane, at 0.1 and 1 million years (m.y.) are calculated to be 0.009 and 0.09 dpa using SRIM2003. However, with the incorporation of 1 wt.% 239Pu, the calculated doses reach 0.27 and ∼1.00 dpa in ten thousand and one hundred thousand years, respectively.

Under oxidizing conditions, multiple cycles of radiation-induced decomposition to UO2 followed by alteration to U6+-phases should be further investigated to determine the fate of trace elements that may have been incorporated in the U6+-phases.

The main conclusions of a Coordinated Research Programmes organized by the International Atomic Energy Agency on the performance of high-level wastes in deep geological repositories are summarized in this paper. The programme ran from 1997 till 2004. Glass, spent fuel and ceramics were the waste forms considered. The 13 participating countries reported their R&D including waste form development, basic understanding of the waste form properties and the performance in simulated disposal conditions. Significant progress has been achieved on the various issues. Recommendations are formulated as to integrate the R&D with the geological disposal conditions and performance assessment considerations, or to conceive generic studies in preparation of the above approach.

Radioactive waste derives from all phases of the nuclear fuel cycle and from the use of radioactive materials in industrial, medical, military and research applications; all such wastes must be managed safely. The most hazardous and long-lived wastes, such as spent nuclear fuel and waste from nuclear fuel reprocessing, must be contained and isolated from humans and the environment for many thousands of years. Many Nuclear Energy Agency (NEA) member countries are, therefore, researching plans for the management of long-lived radioactive waste in engineered facilities, or repositories, located deep underground in suitable geological formations.

A biogeochemical model that represents processes of metal corrosion, microbial degradation of cellulosic waste and mass transfer within a heterogeneous system has been used to represent processes of gas generation in a large-scale (20m3) experiment that has studied degradation of typical nuclear reactor operating waste. The experiment has been in operation for a period of about eight years and has established a pattern of methanogenic gas generation. A “blind testing” approach has been used to develop the model of the experiment using independently derived kinetic data for corrosion and microbial processes. The model correctly represents the anaerobic conditions leading to methane generation during the course of the experiment. The overall rate of gas generation of the experiment is well represented, as is the composition of evolved gases and geochemistry of sampled liquids. The experiment and the model together build confidence in the ability to simulate processes of gas generation and variation in chemical conditions in heterogeneous repository environments.

UK Nirex has supported a programme of work to develop models describing the postclosure evolution of intermediate-level waste packages with the objectives of:

• providing support and justification for the parameters and representations used in performance assessment models;

• informing future model development and packaging advice.

Scenarios for the potential evolution of a waste package were developed and modelled taking explicit account of waste package heterogeneity and the time-dependence of the physical and chemical characteristics of the system. The modelling work highlighted the treatment of organic complexants and the representation of physical containment as two areas in which the impacts of time dependence and package scale heterogeneity might be particularly significant.

A subsequent study of the impact of organic complexants emphasised the importance of heterogeneity in package inventory in determining the radionuclide release from the near field.

The degree of containment afforded by the wasteform and the waste container has been investigated as part of a study to develop a preliminary understanding of the mixing scales within the repository. The study suggests that the most important control on the release of radionuclides from the waste packages is the integrity of the waste encapsulation grout. Interactions between neighbouring packages are to be expected, but the degree to which homogeneous (well mixed) conditions develop may be limited in both time and space.

The extent of the physico-chemical processes of concern in the study on the acceptability of the Belgian Eurobitum bituminised waste for underground disposal will depend on the degree ofageing of the bituminous matrix. Therefore, the ageing of non-radioactive simulates of Eurobitum under gamma irradiation was studied both in the presence and absence of oxygen. Chemical changes in the bitumen structure were measured in the mid-infrared region with the attenuated total reflectance infrared technique (ATR/FTIR) and evaluation of the peaks at 1700 cm−1 (C=O) and 1600 cm−1 (C=C). Even for the highest total absorbed doses (∼800 kGy at ∼150 Gy/h), the effect of the irradiation in Eurobitum was found to be small and was, in any case, smaller than for samples heated for three hours at 130°C in air.

Immobilization of adsorbed Cs+ and Sr2+ on a molybdenum-doped, hexagonal tungsten bronze (HTB)-polyacrylonitrile (PAN) composite adsorbent can be achieved by heating in air at temperatures in the range 600 – 1200 °C. Thermal treatment of the parent composite material at 800 – 1000 °C undergoes a ca. 60% reduction in volume and retains its spherical morphology. For materials prepared at 800 – 1200 °C the full complement of Cs+ and the majority of Sr2+ partition into HTB phases (A∼0.16-0.3MO3; A = Cs, Sr, Na; M = Mo,W), along with sodium cations. The presence of high concentrations of Na+ relative to either Cs+ or Sr2+ does not appear to interfere with the formation of the HTB phase. The fraction (f) of Cs+ and Sr2+ leached from the tungstate phase assemblages is better or comparable with cesium hollandite (Cs0.8Ba0.4Ti8O18; f = ca. 8 × 10−5; rate = <1.2 × 10−4 g m−2d−1) and strontium titanate (SrTiO3; f = 3.1 × 10−3; rate = 2.63 × 10−4 g m−2 day×1), respectively, using a modified PCT test. Furthermore, where aggressive leaching conditions are employed (0.1M HNO3; 150 oC; 4 days), the tungstate phase assemblages display leach resistance orders of magnitude better than the reference phases (Cs+ - f = ca. 5 × 10−3; rate = ca. 1.4 × 10−3 g m−2 day−1; Sr2+ - f = ca. 8 × 10−2; rate = ca. 2.5 × 10−2 g m−2 day−1).

Pursuant to the Decision-in-Principle of 2001 the Finnish programme for geologic disposal of spent fuel has now moved to the phase of underground characterisation of the repository site. The main objective of this programme phase is to confirm – or refute – the suitability of the Olkiluoto site by investigations conducted underground at the actual depth of the repository. The construction work of the access tunnel to the rock characterisation facility (ONKALO) started in the late summer of 2004. The site research and investigations work aims at the maturity needed for submission of the application for construction license of the actual repository in 2012. This requires, however, that also the technology has reached the maturity needed. The design and technical plans form the necessary platform for the development of the safety case for spent fuel disposal. A plan, “road map”, has been produced for the portfolio of reports that should demonstrate the safety of disposal as required by the criteria set by the government and further detailed by the safety authority, STUK.

In Belgium, the Boom Clay is studied as the reference formation for geological disposal of high-level radioactive waste and spent fuel. As the Boom Clay is considered as the main barrier for radionuclide migration/retention, a thorough characterisation of the clay and its pore water was done. This facilitates better understanding of the long-term geological processes and the distribution of the trace elements and radionuclides.

From a mineralogical/geochemical point of view, the Boom Clay is considered as a rather homogeneous sediment, vertically as well as laterally. It is composed of detrital minerals, organic matter and fossils. Minerals are mainly clay minerals, quartz and feldspars. Minor amounts of pyrite and carbonates are also present. Small variations in mineralogical/geochemical composition are related to granulometrical variations. The radiochemical study indicates that the Boom Clay is in a state of secular radioactive equilibrium, meaning that the Boom Clay has not been disturbed for a very long time.

Pore water sampling is done in situ from various piezometers, or by the squeezing or leaching of clay cores in the laboratory. These three pore water sampling techniques have been compared and evaluated. Boom Clay pore water is a NaHCO3 solution of 15 mM, containing 115 mg·1−1 of dissolved natural organic carbon. Some slight variations in pore water composition have been observed and can be explained by principles of chemical equilibrium.