Static aqueous alteration tests were performed with a Japanese simulated HLW glass, P0798, in the presence of bentonite in order to understand the effects of bentonite on the glass alteration kinetics and on the associated Cs release. Analogous alteration tests were performed in 0.001M NaOH solution without bentonite for comparison. The results indicated that; 1) at the initial stage of alteration up to 50 days, no remarkable difference was observed in the alteration rate between both cases “with” and “without” bentonite, 2) at the later stage beyond 50 days, however, the rate in the case “with” bentonite was larger than that in the case “without” bentonite. These results on the alteration rate were analyzed by use of a water-diffusion model. In the case “without” bentonite, a good agreement was observed between the model analysis and the experimental results at the initial stage of alteration up to 50 days, however, the model analysis deviated from the experimental results at the later stage beyond 50 days. In the case “with” bentonite, on the other hand, a good agreement was observed even at the later stage to give the value of the apparent diffusion coefficient, Di of 3.5×10−21m2/s. The comparison between both cases suggests that the alteration rate is controlled by the water diffusion in both cases “with” and “without” bentonite, however, the rate is depressed in the case “without” bentonite probably by the protective effects of the alteration layer developing at the glass surface. In the case “with” bentonite, on the other hand, the alteration layer is expected to be less protective. Cesium desorption tests performed for the altered glass and bentonite indicated that most of the cesium dissolved from the glass is retained in the secondary phase of smectite developing in the precipitated layer by sorption with ion-exchange in the case “without” bentonite. In the case “with” bentonite, however, it is likely to be sorbed at bentonite surface.

Encapsulation of low and intermediate level waste (LLW and ILW) into a monolithic waste form is a key method for preparing such wastes for storage and ultimate disposal. One of the most efficient ways of doing this is by cementation, successfully used for much of today's operational waste in the UK. Formulations are based on composite cements where high levels ofpulverised fuel ash or blast furnace slag replace ordinary Portland cement (OPC). These provide cementing systems with good fluidity, lowered heat of hydration and a matrix with lower porosity and permeability. Further the alkaline hydrated calcium silicate matrix ensures insolubility of many metal hydroxides and hydrated oxides. Dealing with historic or legacy wastes presents a challenge as for many, a suitable cementing formulation has yet to be devised. To achieve ultimate durability, likely interactions between the waste and the cementing matrix need to be well understood when choosing the cement system to be used. The alkaline OPC based matrix causes corrosion with metals such as Al and Mg and porous silicates and layered metal hydroxides used to selectively remove radionucleides react in the alkaline environment, potentially releasing the adsorbed species. Slurries of materials normally considered inert show interactions within the OPC composite systems, although these are not considered detrimental for durability. This paper describes the chemistry of some interactions that occur within OPC composite cements. It also details some results for alternative cementing systems that may provide a toolbox of cement types that may address many of the interaction issues. These include calcium sulphoaluminate cements, inorganic polymers or geopolymers, and activated slag systems.

The Eurobitum bituminised waste product (BWP) from the former EUROCHEMIC reprocessing plant in Mol-Dessel (Belgium) contains a lot of salts, mainly NaNO3 and CaSO4. In contact with pore water in an underground repository, the dehydrated salts in the BWP will rehydrate, resulting in swelling and, possibly, a swelling pressure build-up. A high swelling pressure might impede the integrity and the safety of the repository by creation of preferential pathways for radionuclide migration. In Belgium, the Boom Clay is studied as a reference host formation.

The interaction between the swelling BWP and the host formation is a very complicated hydromechanical process. It depends not only on the hydromechanical behaviour of the Boom Clay, but also on that of the BWP. The hydromechanical constitutive law of Eurobitum is not yet established. We have therefore realised scoping calculations to assess the stress redistribution and deformation due to the recompression of the clay around a disposal gallery caused by a swelling pressure. Sensitivity studies have evidenced some important influencing factors.

The scoping calculations showed that the swelling pressure exercised to the Boom Clay should be limited to 7 - 8 MPa. Beyond this limit, the Boom Clay host formation will be subject to important deformation, and the risk of preferential paths for radionuclide transport becomes very high. The convergence during the excavation of the disposal galleries is of primary importance for a safe waste disposal system design. Indeed, excess convergence may result in a higher damaged zone before disposal of the bituminised waste.

Sorption of selenium (Se) was studied by batch technique by using sedimentary rock samples andcorresponding saline groundwater from Horonobe underground research laboratory (URL)site under reducing condition. Spectroscopic analyses were performed by X-ray absorption near-edge structure (XANES) afterthe sorption experiments to identify the oxidation states of Se on the sedimentary rock. It issuggested thatthe dominant parameters affecting the Se sorption are iron-bearing minerals, especially exchangeable Fe and pyrite, and organic matters. Most of the Se sorbed on the sedimentary rock was Se(0) inthe case of natural saline groundwater, at least at relatively high Se concentration, while only part of the Se was Se(0) in caseof the synthetic groundwater, which contains no organic matters. Incorporationinto pyrite could be the sorption mechanism of Se(0).

Integrated in situ corrosion tests with α-doped SON68 glass samples have been and are still being performed in the HADES underground laboratory in Mol, Belgium. Two test tubes have been dismantled, and the glass and clay samples have been analysed (mass loss, SEM, EDS, SIMS,…). The results on glass corrosion largely respond to the expectations and agree very well with results from surface laboratory tests and modelling predictions. The glass alteration is about two orders of magnitude higher at 90 °C than at 30 °C. The addition of powdered glass frit to a Ca-bentonite backfill reduces the glass alteration by two orders of magnitude in comparison with another backfill, dried Boom Clay. The α-doped glass samples were shown to have slightly thicker alteration layers. The thickness seemed to increase with increasing αβγ-activity. Because of the lack of sufficient data on mass loss, we cannot conclude that also the degree of alteration is higher for these samples. Under the thermal gradient in the test tubes, the reaction of Cabentonite with the glass samples and the powdered glass frit dispersed within the bentonite results in the neo-formation of non-swelling 7 Å minerals.

The structural characteristics of samples in the (Y1−xLax)2Ti2O7 system have been studied as a function of bulk composition using time-of-flight (TOF) powder neutron diffraction, powder X-ray diffraction (XRD), Electron Microscopy and 89Y (I = ½) Magic Angle Spinning Nuclear Magnetic Resonance (MAS NMR). Analysis of the neutron diffraction data suggests the following: For compositions where 0 ≤ x ≤ 0.132, a single cubic phase is present (Fd3, m, Z = 8). Between 0.832 ≤ − ≤ 1, a solid solution with the La2Ti2O7 structure (P21, Z = 4) forms. The two phases coexist between x values of 0.132 and 0.832. XRD suggests that the limits of the two regions of solid solution fall within the ranges of x = 0.1-0.125 and x = 0.875-0.9. The variation in the limits found by the two diffraction methods may be due to stoichiometric errors in the samples. Line-shape analysis of NMR spectra for Y rich compositions suggests that increasing proportions of La are incorporated onto the pyrochlore A site in a statistically random manner, up to the limit of solid solution. The NMR spectra of the monoclinic phase suggest that occupation of the four crystallographically distinct ‘A’ type sites within the structure is not random. At low Y concentrations, two of these sites are preferentially occupied by Y. These sites are suggested to be those found at the edge of the perovskite slabs, which exhibit smaller coordination numbers (based on the number of oxygens within bonding distance of the ‘A’ type cation) than the true perovskite A sites found at the slab centres.

The UK currently uses Portland cement composite blends to immobilise/encapsulate intermediate and low level radioactive wastes (ILW and LLW). However, among other things, the high pH of these systems causes the corrosion of some metals, which can lead to expansion and excess generation of hydrogen. Therefore, in order to immobilise nuclear waste where corrosion is an issue, a near neutral cementing system is desirable.

Among the activators which can be used in the alkali-activated slag (AAS) systems, a solution of Na2SO4 is near neutral but the ground granulated blast-furnace slag (GGBS) itself has a pH about 11, increasing the pH within the Na2SO4 activated AAS. As low calcium Pulverized Fuel Ash (PFA) only has a pH about 9, using a GGBS/PFA blend activated by Na2SO4 offers the potential to develop a near neutral cementing system for nuclear waste immobilisation purposes.

In this paper, the replacement of GGBS in the Na2SO4 activated AAS system with PFA at 0%, 10%, 20% and 30% by mass was examined. The pH and corrosion of Al were determined and used as primary criteria for judging the feasibility for further development of a Na2SO4 activated GGBS/PFA matrix for immobilising nuclear wastes. The microstructure of the matrices was studied by SEM. Leaching studies were carried out to examine the possibility of immobilising Cs+ within these Na2SO4 activated GGBS/PFA matrices. The potential of using Na2SO4 activated GGBS/PFA for immobilising nuclear wastes is discussed.

In order to tackle the particular issues presented by the existence of an Excavation Damaged Zone (EDZ) along disposal cells and tunnels and the subsequent potential occurrence of preferential flow paths and radionuclide migration, a specific Component on the EDZ has been incorporated into the EC Integrated Project on the Near-Field (NF-PRO)1. This Component is organised into several Work Packages matching the foreseeable EDZ evolution during the repository lifetime: (i) the “initial phase”, corresponding to EDZ development induced by underground excavations and rock stress release; (ii) the “transient phase”, corresponding to the repository operating phase, mainly characterised by desaturation, temperature increase and resaturation processes; (iii) the “long-term phase”, corresponding to the post-closure repository period, mainly characterised by EDZ self-sealing / self-healing and gas transport processes.

The scientific results obtained through the numerous tasks included in the Component cover various aspects of EDZ investigation (desk studies, experimental work on representative rock samples, in situ experiments in URLs, numerical modelling) and are integrated into a global EDZ impact evaluation to be forwarded to the Component of the project devoted to process coupling and integration in the general performance assessment of the near-field.

A range of transition metal bearing hollandite phases, formulated Ba1.2B1.2Ti6.8O16 (B2+ = Mg, Co, Ni, Zn, Mn) and Ba1.2B2.4Ti5.6O16 (B3+ = Al, Cr, Fe) were prepared using an alkoxide - nitrate route. X-ray powder diffraction demonstrated the synthesis of single phase materials for all compositions except B = Mn. The processing conditions required to produce > 95 % dense ceramics were determined for all compositions, except B = Mg for which the maximum density obtained was > 93 %. Analysis of transition metal K-edge XANES data confirmed the presence of the targeted transition metal oxidation state for all compositions except B = Mn, where the overall oxidation state was found to be Mn3+. The K-edge EXAFS data of Ba1.2B1.2Ti6.8O16 (B = Ni and Co) were successfully analysed using a crystallographic model of the hollandite structure, with six oxygen atoms present in the first co-ordination shell at a distance of ca. 2.02Å. Analysis of Fe K-edge EXAFS data of Ba1.2B2.4Ti5.4O16 revealed a reduced co-ordination shell of five oxygens at ca. 1.99Å.

Experimental and modelling efforts in the last decade in the frame of nuclear waste management field have been focused on studying the role of the UO2 surfaces in poising the redox state of solid/water systems as well as the radionuclides release behaviour. For this purpose, an experimental programme was developed consisting on dissolution experiments with PWR spent fuel fragments in an anoxic environment and by using different solution compositions.

Some of the collected data has been previously published [1], specifically those data concerning radiolysis products and dissolution of the matrix. The results and the modelling tasks indicated an overall balance of the generated radiolytic species and that uranium dissolution was controlled by the oxidation of the spent fuel matrix in 10mM bicarbonate solutions while in the tests carried out at lower or without carbonate concentrations uranium in the aqueous phase was governed by the precipitation of schoepite.

This paper is the continuation of a series accounting for the data and modelling work related to investigating the release behaviour of minor radionuclides from the spent fuel.

Uranium concentrations as a function of time showed an initial increase until reaching a steady state, indicating a matrix dissolution control. The same behaviour is observed for neptunium, caesium, strontium, technetium and molybdenum indicating a congruent release of these elements with the major component of the fuel matrix. On the other hand, no cler tendency is observed for plutonium data where additional solubility limiting mechanisms may apply.

Kinetic modelling of the trace elements: caesium, strontium, technetium and molybdenum is based on the congruent release of these elements with the major component of the fuel matrix. Rate constants have been determined. Kinetic modelling of neptunium data took also into account the subsequent precipitation as Np(IV) hydroxide. Finally, measured Pu concentrations may be explained by the precipitation of Pu(IV) and/or Pu(III) solid phases.

Among the crystalline matrices being developed for immobilization of 137Cs and 90Sr are framework aluminosilicates like feldspars and feldspathoids. A novel approach to obtain mineral-like aluminosilicate forms of 137Cs and 90Sr with using porous materials based on hollow aluminosilicate microspheres of coal fly ash (cenospheres) has been demonstrated. Two modifications of microspherical porous materials have been developed, (i) moulded porous matrices based on consolidated cenospheres and (ii) zeolite sorbents obtained by hydrothermal zeolitization of cenospheres. The first includes impregnation of Cs- or Sr-containing solutions into porous matrices, drying and calcining at temperatures (700-900°C) lower than the softening point of the porous matrix material, at which solid phase crystallization takes place with formation of framework aluminosilicates, such as Cs-nepheline, pollucite or Sr-anorthite. Using zeolite sorbents, immobilization of Cs+ and Sr2+ proceeds through the step of trapping ions from the solution followed by drying and thermochemical solid-phase transformation of saturated zeolites into pollucite or Sr-anorthite over the same temperature range. Cenosphere-derived porous matrices and zeolite sorbents were tested in fixation of 137Cs using spiked simulants of actual radioactive waste. The durability of crystalline aluminosilicate 137Cs -forms obtained was shown to satisfy Russian (GOST P 50926-96) and international standards.

Coupled modelling has been performed using geochemical/transport codes and radiolysis models to describe the chemical evolution of the waste forms “high-level waste glass” and “spent nuclear fuel” together with its waste package and engineered barrier surroundings. Near field processes considered include container corrosion, hydrogen generation, mass transfer for radionuclides and other waste matrix components in corrosion products and buffer materials, geochemistry of near field solution chemistry, sorption of radionuclides on surface sites in the nano-sized pore space of near field materials and the radiolytic decomposition of pore water. The rate limiting steps in waste form dissolution and secondary phase formation mechanism and the associated radionuclide mobilisation chemistry (solubility, solid solution formation, speciation, redox stability) are strongly influenced by the near field constraints.

Metallic canisters placed in concrete casks and containing spent nuclear fuel elements, will be exposed to a moist oceanic atmosphere while the decay heat generated in the fuel elements cools for more than fifty years. Thus, the surface environment of the metallic canisters will be wet and covered with chloride compounds. The canisters may suffer atmospheric stress corrosion cracking. Therefore, corrosion tests for some potential alloys were conducted in an aqueous bittern solution, containing 22% enriched chloride compounds simulating the expected surface conditions of the canisters, to aid in selecting appropriate alloys based on corrosion resistance. The results suggest that the corrosion resistance of ordinary stainless steels, such as SUS304 and SUS316, is not high enough to avoid ASCC (atmospheric stress corrosion cracking) in the environment. Thus, a higher-grade stainless steel, namely, NSSC270 (20Cr-18Ni-6Mo-0.2N-Low C)or SUS836L (23.5Cr-25Ni-5.5Mo-0.2N-Low C, equivalent to AL-6XN) has been selected for this application.

JNC has developed an uncertainty analysis methodology for application to the spatially heterogeneous characteristics of a geological environment. The developed methodology adopts a new approach that identifies all the possible options in concepts and parameter ranges that cannot be excluded in the light of evidence available. This approach enables uncertainties associated with the understanding at a given stage of the site characterization to be made explicit, using probability theory and possibility theory. The uncertainties could be reduced by screening, to exclude concepts and parameter ranges that can be denied in the light of additional evidence obtained in subsequent investigation stages. This paper describes an outline of the developed methodology and its applicability in the Tono area.

To establish sealing performance suitable for geological environmental conditions of Japan, it is necessary to obtain basic data related for the sealing function through laboratory and in-situ experiments. These data are to be used for practical design and to establish the method for analysis of the sealing performance.

Japan Nuclear Cycle Development Institute (JNC) has joined the international project, the Tunnel Sealing Experiment (TSX), to demonstrate the sealing performance of full scale plugs in-situ, and to develop analytical method of the sealing performance of the plugs at Underground Research Laboratory (URL) of Atomic Energy of Canada Limited (AECL) in Canada [1]. This experiment is in the final stage and fundamental data for the sealing function of clay and concrete plugs have been obtained.

This paper presents the total data handling and interpretation of the sealing experiment, as well as the numerical analysis and interpretation for clay plug performance based on tracer test results. As a result of the numerical analysis, the sealing performance of the clay plug was confirmed and it was shown that the excavation disturbed zones (EDZ) around the clay plug and the tunnel could be a dominant transport pathway for radionuclides.

Cementitious matrices are being used worldwide as a containment medium for radioactive and non-radioactive waste in order to retard the mobility of contaminants. The present thrust of research is to further the understanding of contaminant binding in the cementitious matrix in order to predict the long-term behaviour and the potential impact of the waste on the environment.

The workshop “Mechanisms and Modelling of Waste/Cement Interactions”, held in Meiringen, Switzerland, between May 8 and 12, 2005, focused on the chemical understanding and thermodynamic modelling of the processes responsible for the retention of radioactive and non-radioactive species in cementitious systems. The objectives of the workshop were to bring together scientists from different disciplines, i.e. cement chemistry, radioactive and non-radioactive hazardous waste disposal, to stimulate discussions on current developments and to identify future needs in this field of research. The topics treated in the workshop were chosen to maximize the benefit to the different fields of research. Cement chemists reported on developments in the understanding of cement mineralogy and thermodynamic modelling of cement systems. The hazardous and radioactive waste management communities presented their ideas on the mechanisms of contaminant binding to cement minerals as well as field, laboratory and modelling results from practical applications. In this paper important areas of research on waste/cement interactions presented in the workshop will be outlined and briefly discussed. The following overview reflects a subjective perception of the workshop and does not lay claim to deal comprehensively with all the papers that were presented in the workshop.

Carbon steel overpack will be corroded by consuming oxygen introduced by repository construction after closure of repository and then will keep the reducing environment in the vicinity of repository. The migration of iron corrosion products through the buffer material will affect migration of redox-sensitive radionuclides. Therefore the authors have carried out electromigration experiments with source ofiron ions supplied by anode corrosion of iron coupons in compacted bentonite. However, their migration behavior was complex and difficult to explain. Thus, authors tried to use cesium, whose migration behavior is well known, inthis experimental configuration to obtain knowledge of the migration behavior of cations. The concentrationsof iron and sodium showed nearly complementary distributions. It is expected that iron ion could migrate as ferrous ion through the interlayer of montmorillonite replacing exchangeable sodium ions in the interlayer. On the other hand, cesium profiles seemed to be controlled by ordinary diffusion. Drift of the cesium profile by electric potential gradient could be observed clearly only after 240 h. Apparent dispersion coefficients of cesium were calculated from the profiles and were in reasonable agreement with literature values of apparent diffusion coefficients. Thus this experiment can provide a diffusion field for cations under a reducing condition with ferrous ions in water-saturated bentonite. The effect of electro-osmotic flow on ion migration was negligibly small in this experiment because electro-osmotic flow was compensated by hydraulic pressure caused by the water content gradient developed in the specimen within 24h.

Radionuclide sorption to rock or sediment surfaces in the vicinity of a nuclear waste reposi-tory contributes significantly to the overall safety performance of a given disposal concept. State-of-the-artspeciation techniques as laser - and X-ray spectroscopy are becoming more and more involved into the elucidation of sorption mechanisms and to the quantification of surface speciation. They are extremely valuable tools to gain fundamental understanding as a basis for the development of geochemical sorption models which in turn are needed to predict ra-dionuclide behaviour in the environment.

Nd-bearing zirconolite was leached at 90°C for 157 days in 0.001M citric acid under single-pass-flow-through conditions (modified MCC-4 protocol). Three different flow rates were used, ranging in an order of magnitude from 10 mL per day to 100 mL per day, to determine the effect of the rate of leachant replenishment on the durability of the zirconolite. Results of previous studies on the role of complexing agents on the leaching behaviour of single-phase zirconolite have been included in the discussion.

The pH of the citric acid solution was adjusted to 5 using KOH, mimicking that of the water in the parallel tests, to avoid the influence of pH on chemical durability of the zirconolite.

Simulated groundwater containing 0.001M citric acid at 90°C led to congruency in elemental releases and a diminution of release rate with time of about an order of magnitude, reaching virtual constancy after about 50 to 60 days to a level of about 10−5 g m−2 day−1. The most significant finding was that the elemental release rates of Nd, Ti and Zr (and Ca and Al where detected) were similar for all flow rates. Clearly, varying flow rate by up to an order of magnitude had no effect on elemental releases i.e. there is no solubility limit control on releases at 0.001M citric acid concentration.

An important finding of previous studies using identical leaching protocols with 0.001M citric acid, and inferred in our latest investigations reported here, was that there is no secondary layer development at the surface of the zirconolite to affect leach rates. In contrast, parallel tests carried out in deionised water instead of citric acid showed that hydroxides form in situ on the zirconolite surface, effectively forming hydrolysed zirconolite. This controls further dissolution of the zirconolite matrix due to the solubility limit being reached with respect to the hydrolysed phases rather than with zirconolite. Complexation by citrate ions prevents such control by hydrolysed species on zirconolite solubility.

Even under the more aggressive conditions imposed in these studies (0.001M citric acid), and regardless of flow rate of the leachant, elemental releases from zirconolite are very low for a candidate wasteform and demonstrate its attributes as a ceramic-based wasteform for the containment of actinides.

The radiation damage properties of synthetic pyrochlore-defect fluorite compounds containing lanthanides on the A-site and Ti, Zr, Sn, and Hf on the B-site have been studied extensively using Kr ion irradiation. Using statistical analysis, we show that the results can be quantified in terms of the critical temperature for amorphization, structural parameters, classical Pauling electronegativity difference, and defect energies. The best current model is able to predict the critical temperature to within about 80 degrees Kelvin. The model indicates that radiation tolerance is correlated with an increase in the X anion coordinate toward the value characteristic of the defect fluorite topology, a smaller unit cell dimension, and lower defect energies. Our analysis also demonstrates that radiation tolerance is promoted by an increase in the Pauling cation-anion electronegativity difference or, in other words, an increase in the ionicity of the chemical bonds. Of the two possible cation sites in ideal pyrochlore, the B-site cation appears to play the major role in bonding. This result is supported, for a subset of pyrochlore compounds, by ab initio calculations, which reveal a correlation between the Mulliken overlap populations of the B-site cation and the critical temperature.