Network models of porous media are beneficial for predicting evolution of macroscopic permeability. This work proposes novel models based on truncated octahedral support. Systems with different pore coordination spectra for a given average coordination number can be constructed to match experimental data. This feature, and the allowed pore coordination of 14, make the proposed models more realistic and flexible than existing models with cubic support. Experimental data for two sandstones with substantially different properties are used to demonstrate the models’ ability to predict permeability. A strategy for calculating its evolution with internal damage is also described and results are presented. Developments of this strategy are suggested for deriving mechanism-based constitutive laws for engineering applications.

Batch sorption experiments were carried out, in a glove box under N2 atmosphere, for analysing the effect of increasing ISA (isosaccharinic acid) concentrations on sorption of Tc(IV) and Pu(IV) onto previously characterised Spanish concrete and mortar used in a radioactive waste repository. Results clearly showed that concentrations of ISA around 10-4–10-3 M are able to decrease the distribution coefficients (Rd) of the tetravalent elements analysed, both on concrete and mortar. The decrease in Rd experimentally observed was approximately 2 orders of magnitude for technetium and between 3 and 5 orders of magnitude for plutonium.

SiO2-Al2O3-Y2O3 glasses exhibit high glass transition temperatures, water corrosion resistance and good mechanical properties. These properties suggest that yttrium aluminosilicate glasses could potentially replace the borosilicate glasses usually used for immobilization of nuclear wastes. At the same time, yttrium can be used to simulate actinides.

During waste immobilization, crystallization of the glassy matrix must be avoided or at least controlled, thus, the understanding of glass crystallization kinetics is essential.

We found by XRD that the crystalline phases present on heat treatments are yttrium disilicate and sillimanite/mullite. By optical microscopy on polished cross-sections we could only identify highly yttrium enriched crystals which we associate with yttrium disilicate crystals.

In this paper we measure the surface density of nucleation sites Ns in as obtained splat cooled pieces obtaining values of about 1.5 · 1011 nucleus · m-2. Crystal growth rate U in the temperature range 1000-1040 oC varies in the range 8-13 μm · h-1. These data are useful for designing sintering or melting thermal paths of YAS glasses in order to control their microstructure. We show the effect of glass particle size on DTA results: crystallization peaks moves towards lower temperatures for smaller particle size, which confirms that mainly surface nucleation is taking place on heating.

To perform a safety assessment for the geological disposal of radioactive waste, it is important to understand the response characteristics of the disposal system. In this study, approximate analytical solutions for steady-state nuclide releases from the engineered barrier system (EBS) of a repository were derived for an orthogonal one-dimensional diffusion model. In these approximate analytical solutions, inventory depletion, decay during migration and the influence of groundwater flow in the excavation damaged zone (EDZ) were considered. These solutions were simplified by the Taylor theorem in order to clearly represent the response characteristics of the EBS. The validity of these solutions was shown by comparison with numerical solutions. The response characteristics of the EBS are useful for identifying target values for important parameters that would have the effect of improving the robustness of system safety. The robustness of the geological disposal system and the reliability of the safety assessment can thus potentially be improved using the approximate analytical solutions.

A large amount of liquid radioactive waste has been generated at the Ignalina Nuclear Power Plant (NPP), Lithuania, during its operation. The contaminated liquids are treated with ion exchange-resins, which will generate significant waste stream for cementation. The cemented waste will be disposed of in a near-surface repository. The preliminary safety assessment uncovered that 14C is the most significant radionuclide affecting long-term safety of the closed repository. The method of combined acid striping and wet oxidation with subsequent catalytic combustion has been applied for 14C measurements in cemented ion-exchange resins. It allows separating organic and inorganic compounds from the same sample. At first, the inorganic fraction was extracted by adding acid to the sample followed by absorption of CO2 in a pair of alkali gas washed traps. The remaining carbon was extracted by application of a strong oxidizer. The preliminary results show that activity concentration of 14C in the solidified waste has an order of magnitude of tens and hundreds Bq per gram.

Development of a bioprocess intended to achieve a volume reduction of spent resins (either from research and power reactors) is the main purpose of this research project. Search is constrained to microorganisms that exhibit radioresistance, and which can be cultured in a heavy metal environment with additives of nuclear reactor waters such as boron, lithium and gadolinium compounds.

Bacteria adapted to a radioactive environment were obtained by treating a RA-3 Reactor spent ionic exchange resin sample (kept at Área de Gestión Ezeiza: AGE) with sterile water; microorganisms suspended were isolated, subject to purity controls and characterized.

Experiments performed with these strains include:

-Culture in a mineral-broth having polystyrene as source of both carbon and energy. The strain RMB 1200 gave the highest number of viable cells, then being tested with some aromatic compounds, and with some additives of nuclear reactor waters and other elements found in spent resins.

-An anionic exchange resin sample was heated, yielding a chemical environment resembling to that of radiolysed resins, and then RMB 1200 strain was cultured with this solid.

-A technique to stain polymers was developed in order to find morphological changes caused by microbiological activity. Microscopic examination of cultured polymer samples has been performed.

Results obtained with RMB 1200 include metabolic capabilities to use aromatic carbon sources (benzoate, polyphenols), its growth with polystyrene and resin, and short-term tolerance to several xenobiotics assayed at higher concentration levels than the ones of spent resins.

An overview is given on research activities on cementitious materials for radioactive waste management systems based on the IAEA Coordinated Research Project (CRP) held in 2007-2010. It has been joined by 26 research organizations from 22 countries which shared their research and practical activities on use of cementitious materials for various barrier purposes. The CRP has initially formulated the research topics considered within four specific streams: A) Conventional cementitious systems; B) Novel cementitious materials and technologies; C) Testing and waste acceptance criteria; and D) Modelling long term behaviour.

The CRP has analysed both barrier functions and interactions envisaged between various components with focus on predisposal stage of waste management. Cementation processes have achieved a high degree of acceptance and many processes are now regarded as technically mature. A large body of information is currently available on proven waste conditioning technologies although novel approaches are continuing to be devised.

Most of the existing technologies have been developed for conditioning of large amounts of operational radioactive waste from nuclear power plants and other nuclear fuel cycle facilities. However new waste streams including those resulting from legacy and decommissioning activities required improved material performance and technologies.

The most important outcome of CRP was the exchange of information and research co-operation between different institutions and has contributed towards general enhancement of safety by improving waste management practices and their efficiency. The paper presents the most important results and trends revealed by CRP participants. The research contributions of participating organizations will be published as country contributions in a forthcoming IAEA technical publication.

Alloy 22 is considered as a candidate for engineered barriers of nuclear repositories. Chloride is the only species present in groundwater that is able to promote crevice corrosion, if severe conditions such as high temperatures and a tight crevice are present. Other species present in groundwater have been shown to be inhibitors or non-detrimental species. The objective of this work was to evaluate the efficiency of different species potentially found in groundwaters as possible inhibitors of crevice corrosion of Alloy 22. The crevice corrosion repassivation potential of Alloy 22 was determined in chloride plus inhibitor solutions at 90ºC. The species tested as inhibitors were nitrate, sulfate, carbonate, bicarbonate, chromate, molybdate and tungstate. Nitrate was the most efficient among tested inhibitors. The carbonate was the only species of the carbonate / bicarbonate / carbonic acid equilibrium able to inhibit the chloride-induced crevice corrosion of Alloy 22. Sulfate, chromate and molybdate were moderately good inhibitors.

Degradation of cementitious materials produces leachates of high pH. Such an alkaline plume, if reaching the bentonite buffer, is likely to induce mineralogical and chemical changes in bentonite over long times and may jeopardise the set safety function of the buffer.The objective of this ongoing research is to study the possible alterations of two bentonites, MX-80 and Deponit CA-N, in alkaline leachates at two different temperatures. Also the buffering capacity of the bentonites against high pH will be evaluated.

The ongoing batch experiments are carried out in an anaerobic glove-box (Ar atmosphere, low CO2) at two temperatures (25/60 °C) with three types of simulated cement waters (pH 9.7/9,3, 11.3/10.2 and 12.0/10.9) at 25/0 °C) and one saline groundwater simulate (pH 8.3/7.9) as reference. The solid to liquid ratio used is 1/10. For each set of experiments there are three parallels so that bentonite alteration can be analysed after three different time periods. In the experiment each bentonite sample is leached with several batches of leaching solution. For each renewal of the leaching solution the phases are separated by centrifugation, the reacted solution withdrawn and the chemical composition analysed.

The high-pH experiments (11.3 and 12.0, at 25°C) have continuously shown an initial decrease in the pH-values after each leachate renewal, albeit less dramatic than in the beginning, indicating remaining buffering capacity of the bentonites. The other two experiments (pH 8.3 and 9.7 at 25°C) have shown rather unaltered pH-values. In general, slightly lower pH-values were observed in the Deponit CA-N samples than in those of MX-80. The main cations (Na and Ca) analysed in the leachates have shown a rather expected trends as a result of ion-exchange occurring in the bentonites. The analysed Si concentrations indicate possible dissolution of smectite. More conclusions are possible after the bentonites have been characterized. One experimental set of the 25 °C experiments has been finished and the bentonite phases are being characterized. Other experiment sets are still continued.

Bentonite is planned to be used in many countries as a buffer material in repositories for spent nuclear fuel. The proper understanding and modelling of the functioning of the water-saturated bentonite requires knowledge about the bentonite microstructure and also the way water is distributed between different phases. This paper presents experimental results from our studies of water in compacted, water-saturated MX-80 bentonite at dry densities in the range 0.7-1.6 g/cm3. Three techniques, Cl-porosity, SAXS and proton NMR measurements, were applied to samples kept at room temperature, while TEM imaging was applied to high pressure frozen samples. The combined results of these techniques strongly indicate that the two major water phases in the compacted MX-80 bentonite samples are ‘interlayer’ and ‘non-interlayer’ water. The results of the relative amounts of different water types by SAXS and NMR are very similar. The results by Cl-porosity measurement indicate that only part of the non-interlayer water is available for anions. Those observations are discussed in comparison to TEM micrographs. Our study provides solid experimental evidence for the presence of two major water phases in water-saturated bentonite and estimates their relative proportions and pore sizes.

Since the 1970s there has been a steady increase in research on candidate ceramic and glass-ceramics for immobilisation of HLW and ILW, both from the aspects of crystal-chemical design and processing technology. The variety of ceramics and glass-ceramics designed for different types of HLW and ILW will be presented, notably those which are problematic for vitrification. Several of these materials are optimally processed by hot isostatic pressing (HIP), a technology which can consolidate calcined intermediate-level and high-level nuclear waste. Thus we are targeting such wastes for development of alternative waste forms. The essential process steps during the HIP cycle will be outlined. Effective consolidation of a wide variety of tailored glass-ceramic and ceramic waste forms has been demonstrated. The principal advantages of the HIP technology include negligible offgas during the high temperature consolidation step, relatively small footprint, and high waste/volume loadings. While it can be argued that the “nuclear waste problem” is essentially solved technically, at least with current regulatory guidelines, different perceptions of the “best” waste form and processing method for a given waste, together with the general current lack of agreed locations for final repositories, or even interim storage sites, create uncertainties.

This paper presents a methodology and results on estimating hydraulic properties of the concrete and mortar considered for the near surface disposal facility in Dessel, Belgium, currently in development by ONDRAF/NIRAS. In a first part, we estimated the van parameters for the water retention curve for concrete and mortar obtained by calibration (i.e. inverse modelling) of the van Genuchten model [1] to experimental water retention data [2]. Data consisted of the degree of saturation measured at different values of relative humidity. In the second part, water retention data and data from a capillary suction experiment on concrete and mortar cores was used jointly to successfully determine the van Genuchten retention parameters and the Mualem hydraulic conductivity parameters (including saturated hydraulic conductivity) by inverse modelling.

Cathode reaction on overpack corrosion in the geological repository environments of radioactive waste was identified from corrosion experiments of carbon steel specimens. Carbon steel specimens were encapsulated in degassed glass ampoules with various solutions which were prepared by distilled water degassed by Ar gas bubbling, and set in a thermostatic bath for several weeks. In the X-ray diffraction and X-ray photochemical spectroscopy, crystalline Fe3O4 and Fe2(OH)2CO3, and amorphous Fe(OH)2 were mainly detected on specimens which were immersed into distilled water, high concentration of sodium hydrogen carbonate solution, and low concentration of sodium hydrogen carbonate solution and sodium sulfate solution, respectively. Only hydrogen gas was detected in a gas phase analysis, indicating that hydrogen generation reaction was the dominant cathode reaction in the anoxic condition expected in geological repository environments.

The Finnish spent nuclear fuel disposal is based on the Swedish KBS-3 concept in crystalline bedrock. The concept aims at long-term isolation and containment of spent fuel in copper canisters surrounded by bentonite buffer which mostly consists of montmorillonite. For the long-term modelling of the chemical processes in the buffer, the cation-exchange selectivity coefficients have to be known at different temperatures. In this work, the cation-exchange selectivity coefficients and cation-exchange isotherms were determined in batch experiments for montmorillonite at three different temperatures (25 °C, 50 °C and 75 °C). Five different ratios of NaClO4/Ca(ClO4)2 were used in the experimental solutions. After equilibration the solution and montmorillonite were separated and the solution analysed to get the desired exchange parameters. The experiments were modelled with a computational model capable of taking into account the physicochemical processes that take place in the experiment.

In this paper, I discuss the basis for the following recommendations for the development of standards and regulations for the long-term disposal of spent nuclear fuel and high-level nuclear waste in a geologic repository:

• • The standard and supporting regulations for the licensing of a geologic repository should be generic - applicable to all potential sites. These standards and regulations should be finalized prior to the site-selection process.

• • Site-selection should be based on a set of common-sense criteria [1]. If during site characterization process it is discovered that the site does not meet the technical criteria, the site should be abandoned. These criteria should not only consider the characteristics of the site, but should also include careful consideration of the degree to which a site can be analyzed. Unnecessary complexity may jeopardize the confidence in the analysis of a suitable site.

• • The standard must acknowledge and adapt its structure and standard-of-proof to the fact that there are two time-scales of interest: the human time-scale that extends to some thousands of years and the geologic time-scale that extends to many hundreds of thousands of years. Reasonable and robust containment at both time scales is possible, but the type of analysis and standard-of-proof will be different for each.

• • Because there are two time-scales and because the types of “proof” for each are very different, the total system analysis of performance, reduced to a single numerical estimate of risk at some very distant time, should be abandoned. The standard should not require scientists and engineers to complete an analysis that is at its best opaque and at its worst not believable.

Alloy 22 has been extensively studied regarding its crevice corrosion (CC) resistance both in pure chloride solutions and in solutions containing different oxyanions that may act as inhibitors of crevice corrosion. The scope of this work was to study the general and localized corrosion behavior of Alloy 22 when phosphate ions were added to a 1 M NaCl solution at 90°C. Results from the electrochemical tests indicate that the size of the passive potential range and the localized corrosion repassivation potential value increased in the presence of phosphate ions. Results from creviced specimens showed a strong inhibition effect of phosphate ions on the chloride induced crevice corrosion of Alloy 22. The critical molar concentration ratio (RCRIT = [phosphate]/[Cl]) to inhibit crevice corrosion was 0.3.

ThermoChimie is the thermodynamic data base initially created by ANDRA in 1996, especially designed and qualified for systems of interest for the French high level waste repository concept. This database is supported by an experimental program on actinides and fission products and also on major components of the systems of interest, and it has been continuously updated since its creation. The validation of the database is continuously on-going through geochemical calculations related to the performance assessment of different backfill/buffer materials and/or geological formations. ThermoChimie contains data on major elements (including stability of minerals such as clays, zeolites, cementitious phases), a long list of radioelements, such as actinides and lanthanides, chemotoxic metals, as well as organic and inorganic ligands.

All the countries that operate commercial nuclear power plants are planning to dispose of the waste in underground geologically stable repositories. The materials being studied for the fabrication of the containers include carbon steel, stainless steel, copper, titanium and nickel alloys. The aim of this work is to review results from research performed using the alloys of interest regarding their resistance to environmentally assisted cracking (EAC) under simulated repository conditions. In general, it is concluded that the environments are mild and that the studied metals may not be susceptible to cracking under the planned emplacement conditions.

In recent years, nuclear waste management has become a fundamental issue in the nuclear energy production cycle. Tomographic Gamma Scanner (TGS) is an essential tool for nuclear waste characterization. It is crucial to rely on local support and cost effective solutions; for this reasons, we are designing our TGS system based on local technology. In this work, we present a study of different geometries and instrumentation chain parameters to design a TGS.

A set of Monte Carlo simulations were performed to evaluate energy and spatial resolution limitations of scintillator, CZT (Cadmium Zinc Telluride), and HPGe (high purity germanium) detectors. Collimator and detector geometries were studied to maximize the characteristics of the system. In this study, a phantom of 137Cs and 60Co was utilized to evaluate the overall performance of the proposed TGS system. In addition, the impact of electronic instrumentation chain and image reconstruction algorithms was taken into account.