The possibility of a simple heating process of POM to obtain tungsten bronze was investigated for nuclear waste immobilisation via DTA/TG and high temperature XRD. Heating process up to 900°C caused the decomposition of structure for both systems. Cooling process seemed to have little effect on the final product for the K11[Nd(PW11O39)2]·xH2O, whereas the cooling profile showed a significant effect on the K13[Nd(SiW11O39)2]·xH2O. Nd formed two types of tungsten bronzes, namely Nd2WO6 and Nd4W3O15 in K11[Nd(PW11O39)2]·xH2O and K13[Nd(SiW11O39)2]·xH2O, respectively.

The progress in study of immobilization of liquid radioactive waste in mineral-like aluminosilicate matrices (feldspar and feldspathoids) derived from cenosphere-based microspherical zeolite sorbents is reported. Based on the analysis of ‘composition – morphology –C characteristics’ correlation obtained for more than 70 cenosphere products with a high content of one morphological type of imperforated cenospheres, optimal ranges of cenospheres composition were identified for the synthesis of zeolites with the predetermined structural type. Given that the cenosphere-based zeolite sorbents are precursors of the final waste form, further activity proceeds toward developing a highly efficient zeolite sorbent of NaP1 type, which would ensure the effective trapping of Cs+ and/or Sr2+ ions and the phase transformation under relatively mild conditions (T≤1000°C). The paper details the main criteria for realization of the ‘precursor’ approach for fixation of Cs-Sr fraction using cenosphere-based systems; also shown are experimental results indicating practical realization of the proposed approach.

Bulk samples of six Ln2TiO5 compounds with Ln = La, Pr, Nd, Eu, Gd and Tb were prepared and characterised. Most of the samples have a phase purity of ∼95% (based on BEI and EDS) with the predominant secondary phase primarily being Ln2Ti2O7. Using XRD, TEM selected area diffraction and high resolution imaging techniques, we have confirmed the results of previous studies which showed that at room temperature Pr2TiO5, Nd2TiO5, Eu2TiO5 and Tb2TiO5 have orthorhombic structures with Pnma symmetry. The structure of Tb2TiO5 was further monitored as a function of temperature. The relevance of Ln2TiO5 compounds to advanced nuclear fuel cycles is discussed.

In 1987 samples of Pu238 and Cm244 doped Synroc C were prepared in the UKAEA Laboratories at Harwell. They were studied for five years before being archived. During decommissioning of the Harwell laboratories the samples were transferred to Sellafield and the opportunity was taken to conduct further studies. Given the age of the samples, they offer a unique insight into the long term radiation stability of Synroc. To date, the three Pu238 samples have been examined. The alpha decay dose experienced by the samples is estimated to be 3 × 1019 alphas per gram.

The sample allowed to accumulate alpha decay damage (10587) was slightly heterogeneous, with an apparent grain size in the range 3–15νm. EDX analysis confirmed the phases present to be those expected in Synroc C and that the Pu had partitioned predominantly into the perovskite and zirconolite. Microcracking was observed in the hollandite and rutile but cracks arrested once they reached a zirconolite or perovskite grain.

Sample 10588 was annealed after five years, at temperatures up to 1200°C, this sample was microstructurally similar to 10587 at the 20-50νm scale but differs at small scales. A major difference is the presence of both intergranular and intragranular porosity.

Sample 10589 was annealed at the same time as 10588 but differed significantly probably due to actual fabrication temperatures being higher than those recorded.

Information on these samples will be presented, along with a discussion of the implications for the expected long term stability of titanate ceramic wasteforms.

To investigate the resistance of actinide host phases to accelerated radiation damage, which simulates radiation induced effects of long term storage, the following samples doped with plutonium-238 (from 2 to 10 wt. %) have been repeatedly studied using XRD and other methods: cubic zirconia, Zr0.79Gd0.14Pu0.07O1.99; monazite, (La,Pu)PO4; ceramic based on Pu-phosphate of monazite structure, PuPO4; ceramic based on zircon, (Zr,Pu)SiO4, and minor phase tetragonal zirconia, (Zr,Pu)O2; single crystal zircon, (Zr,Pu)SiO4; single crystal monazite, (Eu,Pu)PO4; ceramic based on Ti-pyrochlore, (Ca,Gd,Hf,Pu,U)2Ti2O7. No change of phase composition, matrix swelling, or cracking in cubic zirconia were observed after cumulative dose 2.77×1025 alpha decay/m3. The La-monazite remained crystalline at cumulative dose 1.19×1025 alpha decay/m3, although Pu-phosphate of monazite structure became nearly amorphous at relatively low dose 4.2×1024 alpha decay/m3. Zircon has lost crystalline structures under self-irradiation at dose (1.3-1.5)×1025 alpha decay/m3, however, amorphous zircon characterized with high chemical durability. The Ti-pyrochlore after cumulative dose (1.1-1.3)×1025 alpha decay/m3 became amorphous and lost chemical durability. Radiation damage caused crack formation in zircon single crystals but not in the matrix of polycrystalline zircon. Essential swelling and crack formation as a result of radiation damage were observed in ceramics based on Ti-pyrochlore and Pu-phosphate of monazite structure, but not so far in La-monazite doped with 238Pu.

Samples of murataite ceramics with the composition (wt.%) 3.8 Al2O3, 10.5 CaO, 54.0 TiO2, 10.6 MnO, 6.0 Fe2O3, 4.6 ZrO2, 8.1 ThO2, 2.4 Cm2O3 (1.8 244Cm) and a specific activity of 5.5×1010 Bq/g were prepared by cold pressing and sintering at 1250 °C for 24 hrs or by melting and recrystallisation in a resistive furnace at 1325 °C and 1350 °C for 1 hr. In the sintered ceramics murataite polytypes with five-fold (5C) or three-fold (3C) repeats of the fluorite unit cell and crichtonite were found to be the major phases. Perovskite, pseudobrookite, and pyrochlore were observed as minor phases. The 5C polytype was rendered X-ray amorphous at a cumulative dose of 2.73×1018 μ–decays/g (0.21 dpa) whilst the 3C polytype, which contained only traces of Cm, remains crystalline at this dose. In the melted ceramics the 5C and 8C murataite polytypes were found to be the major phases (80–90 % of the bulk) and minor amounts of rutile, crichtonite and perovskite were also observed. Complete amorphization of the murataite polytypes in the ceramics melted at 1325 and 1350 °C was achieved at doses of 2.46×1018 μ-decays/g (0.19 dpa) and 2.53×1018 μ-decays/g (0.20 dpa), respectively.

A vitrification technology utilizing a lanthanide borosilicate (LaBS) glass is a viable option for dispositioning excess weapons-useable plutonium that is not suitable for processing into mixed oxide (MOX) fuel. A significant effort to develop a glass formulation and vitrification process to immobilize plutonium was completed in the mid-1990s. The LaBS glass formulation was found to be capable of immobilizing in excess of 10 wt % Pu and to be tolerant of a range of impurities. A more detailed study is now needed to quantify the ability of the glass to accommodate the anticipated impurities associated with the Pu feeds now slated for disposition.

The database of Pu feeds was reviewed to identify impurity species and concentration ranges for these impurities. Based on this review, a statistically designed test matrix of glass compositions was developed to evaluate the ability of the LaBS glass to accommodate the impurities. Sixty surrogate LaBS glass compositions were prepared in accordance with the statistically designed test matrix. The heterogeneity (e.g. degree of crystallinity) and durability (as measured by the Product Consistency Test – Method A (PCT–A)) of the glasses were used to assess the effects of impurities on glass quality.

The UK inventory of separated civil plutonium is expected to exceed 100 tonnes by 2010. Whilst the majority of this could be used in the manufacture of MOx (Mixed Oxide) fuel in future power generation scenarios, options for the disposal of surplus plutonium are currently being investigated by Nexia Solutions Ltd on behalf of the UK's Nuclear Decommissioning Authority (NDA). One of the options being considered is immobilisation in a durable glass matrix followed by long term storage and subsequent final repository disposal.

A preliminary experimental survey assessed a selection of potential glass systems on the basis of Pu-surrogate (cerium) loading, durability, and ease of processing. Following this, a number of borosilicate compositions have been taken forward into a more detailed investigation in order to fully qualify their potential for Pu-immobilisation. The selected compositions are lanthanide borosilicate (LaBS), alkali tin silicate (ATS) and high-lanthanide alkali borosilicate (modified-MW). For this second series of experiments, hafnium was selected as the Pu surrogate, and a study of the potential waste loading as a function of temperature for the three selected compositions is described in this paper. Furthermore, several variations of the LaBS composition were fabricated in order to investigate the effect of total lanthanide content on melting temperature. The benchmark of 10 wt% HfO2 incorporation is achievable for all three glasses with temperatures of 1200, 1300 and 1400 °C required for ATS, modified-MW and LaBS respectively.

The UK has a considerable stockpile of separated plutonium; a legacy of over 50 years of civilian nuclear programmes. This material has been considered both as an asset for future energy generation and a liability due to the proliferation threat. A proportion of the PuO2 stocks may be consumed by nuclear fission, in mixed oxide (MOx) or inert matrix (IMF) fuels but a quantity of waste PuO2 will remain which is unsuitable for fuel manufacture and will require immobilisation. A research program is currently underway to investigate the potential of various single phase ceramic formulations for the immobilisation of this waste PuO2 fraction. In this work a number of synthetic mineral systems have been considered including titanate, zirconate, phosphate and silicate based matrices. Although a wealth of information on plutonium disposition in some of the systems exists in the literature, the data is not always directly comparable which hinders comparison between different ceramic hosts. The crux of this research has been to compile a database of information on the proposed hosts to allow impartial comparison of the relative merits and shortcomings in each system.

Surrogates are widely used in the research and development of nuclear wasteforms, providing detailed insight into the chemical and physical behaviour of the wasteform whilst avoiding the widespread (restricted and costly) use of radiotoxic elements in the laboratory. However, caution must be exercised when dealing with surrogates since no single element or compound perfectly mimics all aspects of the behaviour of another. In this paper we present a broad discussion of the use of surrogates in waste immobilization, drawing upon and highlighting our research into glass and ceramic wasteforms for the immobilization of bulk PuO2.

Several glass compositions are currently under investigation for immobilisation of the separated PuO2 that has been produced as a result of civil nuclear fuel reprocessing in the UK. Whilst a final decision on the fate of what ultimately will be over 100 tonnes of plutonium has yet to be made, all options for the disposition of this material are currently being investigated by Nexia Solutions Ltd on behalf of the Nuclear Decommissioning Authority (NDA).

As one of the immobilisation options, vitrification in borosilicate glass could potentially provide a criticality-safe and stable waste form with durability suitable for long term storage and subsequent repository disposal. From an initial experimental survey of potential candidates, three borosilicate compositions were selected for a more detailed study of the waste loading and chemical durability: lanthanide borosilicate (LaBS), alkali tin silicate (ATS) and high-lanthanide alkali borosilicate (modified-MW). In these inactive tests, hafnium was used as the surrogate for plutonium. This paper describes a range of static leach tests that were undertaken in order to understand the overall durability of the waste forms, as well as the release rates of the Pusurrogate when compared to any neutrons poisons present in the glass. For the LaBS compositions it was found that the release rate of gadolinium was potentially slightly higher than that of hafnium, although both were as low as 10-5 to 10-6 g m2 day-1. The potential implications for long-term repository behaviour are discussed.

The so-called Instant Release Fraction (IRF) is considered to govern the dose released from Spent Fuel repositories. Often, IRF calculations are based on estimations of fractions of inventory release based in fission gas release [1]. The IRF definition includes the inventory located within the Gap although a conservative approach also includes both the Grain Boundary (GB) and the pores of restructured HBS inventories.

A correction factor to estimate the fraction of Grain Boundary accessible for leaching has been determined and applied to spent fuel static leaching experiments carried out in the ITU Hot Cell facilities [2]. Experimental work focuses especially on the different properties of both the external rim area (containing the High Burn-up Structure (HBS)) and the internal area, to which we will refer as Out and Core sample, respectively. Maximal release will correspond to an extrapolation to simulate that all grain boundaries or pores are open and in contact with solution.

The correction factor has been determined from SEM studies taking into account the number of particles with HBS in Out sample, the porosity of HBS particles, and the amount of transgranular fractures during sample preparation.

The so-called Instantaneous Release Fraction (or IRF) corresponds to the rapid release of radioactivity by the spent fuel rod at the canister breaching time in repository and depends on the spent fuel evolution in a closed system. The effect of He accumulation on the spent fuel pellet microstructure and diffusion processes are here re-assessed in the light of the recent results issued from the projects Near-Field PROcesses and PRECCI. It allows diminishing the conservatism of the former IRF values of key safety relevant radionuclides for PWR UO2 fuels.

Np-237 (τ1/2 = 2.1 million years) is a potentially important contributor to the total dose for a geologic repository under oxidizing conditions. Further, the Np5+-complexes are mobile aqueous species. Several processes may limit the transport of Np, as well as other actinides: i) the precipitation of Np-solids, ii) the incorporation of Np into secondary uranium phases, and iii) the sorption and reduction of Np-complexes on Fe-oxide surfaces. This study utilizes quantum-mechanical calculations to determine the most energetically favorable Np5+-incorporation mechanisms into uranyl phases, where Np5+-substitution for U6+ requires a charge-balancing mechanism, such as the addition of H+ into the structure. Experimental results suggest that uranyl structures with charged interlayer cations have a greater affinity for Np5+ than uranyl structures without interlayer cations. Therefore, the uranyl silicate phase boltwoodite (KUO2(SiO3OH)(H2O)1.5) is selected for this computational investigation. The charge-balancing mechanisms considered to occur with substitution include: i) addition of H+, ii) substitution of Ca2+ for K+, and iii) substitution of P5+ for Si4+. While the incorporation energy results (1-3 eV)are higher than energies expected based on current experimental studies, solid-solution calculations are used to estimate the limit of Np incorporation for the P5+ substitution mechanism (10 ppm at ̃100°C). The electronic structure of the boltwoodite structure provides insight into the electron density that may be involved in the incorporation of Np into the structure.

The immobilization of U(VI) by C-S-H phases under conditions relevant for the cementitious near field of a repository for radioactive waste has been investigated. C-S-H phases have been synthesized using two different procedures: the “direct reaction” method and the “solution reaction” method.

The stabilities of alkaline solutions of U(VI) (presence of precipitates or colloidal material) were studied prior to sorption and co-precipitation tests in order to determine the experimental U(VI) solubility limits. These U(VI) solubility limits were compared with the U(VI) solubilities obtained from thermodynamic speciation calculations assuming the presence of combinations of different solid U(VI) phases. The solid phase controlling U(VI) solubility in the present experiments was found to be CaUO4(s).

The U(VI) uptake kinetics and sorption isotherms on C-S-H phases with different C:S ratios were determined under various chemical conditions; e.g., sorption and co-precipitation experiments and different pH’s. U(VI) was found to sorb fast and very strongly on C-S-H phases with distribution ratios (Rd values) ranging in value between 103 L kg-1 and 106 L kg-1. Both sorption and co-precipitation experiments resulted in Rd values which were very similar, thus indicating that no additional sorption sites for U(VI) were generated in the co-precipitation process. Furthermore, C-S-H synthesis procedures did not have a significant influence on U(VI) uptake. The U(VI) sorption isotherms were found to be non-linear, and further, increasing Ca concentrations resulted in increasing U(VI) uptake. The latter observation suggests that U(VI) uptake is controlled by a solubility-limiting process, while the former observation further indicates that pure Ca-uranate is not the solubility-limiting phase. It is proposed that a solid solution containing Ca and could control U(VI) uptake by C-S-H phases.

In this study, Ca, Mg, Al, and Pb concentrations leached from uncarbonated and carbonated ordinary Portland cement – dried waste incinerator bottom ash samples during single extraction tests (EN12457 test) at a pH from 1 to 12, were modelled using the geochemical code PHREEQC. A good agreement was found between modelling results and experiments in terms of leached concentrations for Ca, Mg, and Al by defining a single set of pure mineralogical phases for both the uncarbonated and carbonated (three levels) samples. The model also predicted well the observed decrease in Ca leaching with increasing carbonation. Modelling results further revealed that leaching of Pb is not controlled by dissolution/precipitation of pure Pb containing minerals only (carbonates and (hydr)oxides). The addition of solid solutions (calcite-cerrusite and gibbsite-ferrihydrite-litharge solid solutions) and adsorption reactions on amorphous Fe- and Al-oxides improved the model representation of the experimentally observed amphoteric leaching profile of Pb from the cementitious material.

Two infiltration tests with a synthetic hyperalkaline solution at pH 13.5 were performed. The aim was to analyse the interaction of concrete solutions with compacted smectitic bentonites. The experiment was performed inside an anoxic glove box at 30-35 °C with FEBEX bentonite compacted at a dry density of 1.65 g/cm3 and a water content of 13.4%. Outflowing water coming from the bentonite was collected over time. A total of three times the bentonite pore volume was recovered and twelve chemical analyses of the resulting pore water were obtained. The ionic strength of the pore waters decreases with time from 0.20 M to 0.03 M and the pH increases from 7.9 to 9.0. After a test period of 595 days, the cells were dismantled, and the study of the state and alteration of the bentonite was undertaken. A modification in the a sorbed cation population, an increase in the Na- and K-smectites type, and the presence of a tri-octahedral smectite (saponite) and zeolites (phillipsite and chabazite) were observed.

Bentonite plays a significant barrier role in many radioactive waste repository designs, where it has been chosen due to its favourable properties such as plasticity, swelling capacity, colloid filtration, low hydraulic conductivity and its stability in relevant geological environments. However, bentonite is unstable at high pH meaning that it could lose its favourable properties if interacted with hyperalkaline leachates from concrete construction materials (e.g. tunnel liners, grouts, etc.), seals and plugs and/or cementitious wastes in a repository. This fact has forced several national programmes to assess alternative construction and sealing materials such as low alkali cements. Recently, it has been assumed that the lower pH (typically pH 10-11) leachates of such cements will degrade bentonite to a much lesser degree than ‘standard’ OPC-based cement leachates (generally with an initial pH>13).

To date, few laboratory or in situ URL (underground rock laboratory) data are available to support the use of low alkali cements in conjunction with bentonites, partly because of the very slow kinetics involved. Consequently, a new project has focussed on finding an appropriate natural analogue site to provide long-term supporting data which will avoid the kinetic constraints of laboratory and URL experiments. Early results have identified an initial, very promising site at Mangatarem in the Philippines, where a quarry excavating bentonite and zeolites is found in the sedimentary carapace of the Zambales ophiolite. In the immediate vicinity of the quarry, ophiolite-derived hyperalkaline groundwaters are present and further field work (including geophysics surveys and borehole drilling) are now being planned to assess regional bentonite/hyperalkaline groundwater interaction. This paper presents an overview of the current status of the project and assesses the relevance of the study to improving understanding of low-alkali cement leachate/bentonite interaction.

Carbon steel overpack is corroded by consuming oxygen introduced by repository construction after closure of the repository and then maintains the reducing environment in the vicinity of the repository. The migration of iron corrosion products through the buffer material will affect the migration of redox-sensitive radionuclides. Therefore, it is important to study the migration of iron corrosion products through the buffer material because it may affect the corrosion rate of overpack, and migration of redox-sensitive radionuclides. Electromigration experiments have been conducted with the source of iron ions supplied by anode corrosion of the iron coupon in compacted bentonite. The carbon steel coupon was connected as the working electrode to the potentiostat and was held at a constant supplied potential between - 650 to +300 mV vs. Ag/AgCl electrode for up to 168 hours. The amount of iron penetrated into a bentonite specimen was in good agreement with the calculated value from the corrosion current under the assumption that iron is dissolved as ferrous ions. A model using dispersion and electromigration could explain the measured iron profiles in the bentonite specimens. The fitted value of electromigration velocity depended on the potential supplied. On the other hand the fitted value of the dispersion coefficient did not depend on the potential supplied but a constant. This constant dispersion coefficient could be due to the much larger diffusion coefficient of ferrous ion in bentonite compared with the effect of mechanical dispersion. The experimental configurations used in this study are applicable to the examination of the migration behaviour of cations with the source of iron ions under a reducing condition controlled with a potentiostat.

Electrochemical studies such as cyclic potentiodynamic polarization (CPP) and electrochemical impedance spectroscopy (EIS) were performed to determine the corrosion behavior of Alloy 22 (N06022) in 1M NaCl solutions at various pH values from acidic to neutral at 90°C. All the tested material was wrought Mill Annealed (MA). Tests were also performed in NaCl solutions containing weak organic acids such as oxalic, acetic, citric and picric acids.

Results show that the CR of Alloy 22 was significantly higher in solutions containing oxalic acid than in solutions of pure NaCl at the same pH. Citric and Picric acids showed a slightly higher CR, and Acetic acid maintained the CR of pure chloride solutions at the same pH. Organic acids revealed to be weak inhibitors for crevice corrosion. Higher concentration ratios, compared to nitrate ions, were needed to completely inhibit crevice corrosion in chloride solutions.

Results are discussed considering acid dissociation constants, buffer capacity and complex formation constants of the different weak acids.