Large volumes of depleted natural and low-enriched uranium exist in the UK waste inventory. This work reports on initial investigations of the leaching performance of candidate glass and cement encapsulation matrices containing UO3 powder as well as that of uranium oxide powders. The surface areas of UO3 powder and the monolith samples of UO3 conditioned in the glass and cement matrices were very different making leaching comparisons difficult. The results showed that for both types of monolith conditioned samples a steady increase of uranium concentration in solution with time was generally not observed. The wt.% of uranium leached from UO3 conditioned in the lead borosilicate glass wasteform was approximately five orders of magnitude less than that leached from UO3 powder. Similarly, the quantities of uranium leached from UO3 conditioned in composite cement made with ordinary Portland cement, and from magnesium phosphate cement, were approximately four and three orders of magnitude, respectively, less than that leached from UO3 powder. The performance of a mixed oxide borosilicate glass wasteform was only slightly better than that of UO3 powder. This work shows that wasteforms based on encapsulation in lead borosilicate glass and cement matrices have the greatest potential for further development.

With potential storage of several hundred years underground in a geological disposal facility (GDF) before closure, there is a requirement for radioactive waste packages to perform adequately. Over the past 25 years, cementitous grouts based on blast furnace slag (BFS) and ordinary Portland cement (OPC), have been used in the UK to immobilize intermediate-level waste (ILW), Magnox swarf, and currently such wasteforms are in surface storage awaiting geological disposal. Magnox fuel cladding will slowly corrode when encapsulated in alkaline cementitious grouts to produce hydrogen gas and an expansive corrosion product. Expansive corrosion products may lead to degradation of the wasteform and, if extensive, could affect the container. This study investigated the acute and chronic rates of corrosion of unirradiated Magnox swarf encapsulated in a BFS/OPC grout over a range of temperatures (25, 40, 60, 75 and 90°C) for curing times of up to 2½ years. Structural product degradation starts to develop as the tensile strength of the grout is approached by the expansive forces generated. Deformation of some experimental containers was also noted. An estimate of the time taken for a grouted product to fracture from this study due to the corrosion solely of Magnox is 350 years for storage at 25°C. Although not fully described in this paper, the main cement phases were calcium silicate hydrate, portlandite and gehlenite; the main product of Magnox corrosion was brucite. The curing temperature did not affect the compositions of the Magnox metal, corrosion product or grout.

Vitrification is used for the immobilization and conditioning of high-level waste (HLW) arising from the reprocessing of spent nuclear fuel in the UK. Vitrification is also under consideration for the immobilization of certain intermediate-level wastes (ILW), where there may be advantages of volume reduction and removal of uncertainties in long-term waste behaviour, compared to encapsulation in a cement grout. This paper gives an overview of recent work into the chemical durability of UK vitrified wasteforms to inform the technical specification for the disposal facilities for these waste products and the treatment of their long-term behaviour in post-closure performance assessment. This has included: (1) measurements of the initial glass dissolution rates of a simulated HLW Magnox waste glass in a range of groundwater types representative of potential UK host geologies and in simulated high pH near-field porewaters relevant to potential disposal concepts, using Product Consistency Test type-B (PCT-B) at 40°C; and (2) durability testing of three simulant ILW glasses in a saturated calcium hydroxide buffered solution to simulate conditions in cement-based disposal vaults, using PCT-B tests at 50°C.

The experimentally defined initial rate of HLW Magnox waste glass dissolution in a range of simulated groundwater compositions appears to be similar regardless of the ionic strength and major element composition of the solution. The release of caesium from HLW Magnox waste glass appears to be sensitive to solution composition. Caesium is selectively retained in the glass compared to other soluble components in the two low ionic strength solutions, but is released at similar rates to other soluble components in the three groundwaters and Ca(OH)2 solution. Whether this change in caesium retention is an ionic strength effect or is related to changes in the nature of the surface alteration layer formed on the glass, has yet to be established. For HLW Magnox waste glass, dissolution is accelerated at high pH in NaOH solution, however, the presence of calcium acts to mitigate the effects of high pH, at least initially. In Ca(OH)2 solution, calcium is found to react with all the glasses studied leading to the formation of calcium-containing alteration products. The initial dissolution behaviour in Ca(OH)2 solution varies with glass composition and in particular appears to be sensitive to the boron content.

In the Swedish KBS-3 repository concept, there is potential for gas to be generated from corrosion of ferrous materials under anoxic conditions, combined with the radioactive decay of the waste and radiolysis of water. A full understanding of the probable behaviour of this gas phase within the engineered barrier system (EBS) is therefore required for performance assessment. We demonstrate key features from gas transport experiments on pre-compacted Mx80 bentonite, under laboratory and field conditions, and discuss their implications in terms of a conceptual model for gas migration behaviour. On both scales, major gas entry is seen to occur close to the sum of the porewater and swelling pressures of the bentonite. In addition, gas pressure at breakthrough is profoundly sensitive to the number and location of available sinks for gas escape. Observations of breakthrough can be explained by the creation of dilatational pathways, resulting in localized changes in the monitored porewater pressures and total stresses. These pathways are highly unstable, evolving spatially and temporally, and must consequently influence the gas permeability as their distribution/geometry develops.

Such observations are poorly embodied by conventional concepts of two-phase flow, which do not fully represent the key processes involved. Although dilatancy based models provide a better description of these processes, the paucity of data limits further development and validation of these models at present.

In the UK, radioactive wastes currently planned for disposal in a geological disposal facility (GDF) include intermediate level waste, some low level waste and high level waste. Disposal of other materials, including spent fuel, uranium and plutonium is also being evaluated to inform the safety case for a GDF, if such materials were to be classified as wastes in the future. This paper describes the generic safety functions through which waste packages can contribute to the safety case of a GDF in the UK. It describes the engineering approach used or envisaged, in the UK and internationally, to ensure that waste packages retain their safety functions for the required periods of time and summarizes the scientific basis underpinning the current understanding of relevant evolution processes. Where gaps in the knowledge exist, the Nuclear Decommissioning Authority Radioactive Waste Management Directorate has identified specific research activities needed to close out such gaps to a level of maturity sufficient for this stage of the disposal programme (generic). This paper describes the latest results from their R&D programme and presents a summary of the research activities planned to meet the current needs of the disposal programme with specific reference to the topics of package evolution and criticality safety.

Pore-volume changes in porous media during water–rock interaction can be studied using hydrological tracers. The tracers used here were amino G acid, napthionic acid and fluorescein at pH 3 and 6.5 in contact with basaltic glass, quartz and rhyolite. The experimental setup mimicked that of a hydrological tracer test where a fixed volume of tracer was injected into a flow-through column and the breakthrough curve monitored. The measured breakthrough tracer curves were compared to theoretical 1-D reactive transport simulations calculated using the PHREEQC program. In some cases the tracers were observed to behave ideally, whereas in others they clearly reacted with the solid surfaces. This implies that some common hydrological tracers used in groundwater hydrology may not be suitable under all conditions as they may react with the surrounding rocks in the groundwater system.

Natural hydromagnesite (Mg5(CO3)4(OH)2·4H2O) dissolution and precipitation experiments were performed in closed-system reactors as a function of temperature from 22.5 to 75ºC and at 8.6 < pH < 10.7. The equilibrium constants for the reaction Mg5(CO3)4(OH)2·4H2O + 6H+ = 5Mg2+ + 4HCO3– + 6H2O were determined by bracketing the final fluid compositions obtained from the dissolution and precipitation experiments. The resulting constants were found to be 1033.7±0.9, 1030.5±0.5 and 1026.5±0.5 at 22.5, 50 and 75ºC, respectively. Whereas dissolution rates were too fast to be determined from the experiments, precipitation rates were slower and quantified. The resulting BET surface areanormalized hydromagnesite precipitation rates increase by a factor of ~2 with pH decreasing from 10.7 to 8.6. Measured rates are approximately two orders of magnitude faster than corresponding forsterite dissolution rates, suggesting that the overall rates of the low-temperature carbonation of olivine are controlled by the relatively sluggish dissolution of the magnesium silicate mineral.

The release and migration of gaseous carbon-14 has been identified as a key issue for geological disposal of intermediate-level radioactive wastes in the UK. A significant fraction of carbon-14 in the UK inventory is in irradiated graphite. This paper describes measurements of gaseous carbon-14 releases from irradiated graphite on immersion in alkaline solution. Apparatus has been developed to discriminate organic and inorganic (14CO/14CO2) species in the gas phase by means of selective oxidation and capture. In the initial experiment, small amounts of gaseous carbon-14 (∼4 Bq) were released from 9 g of crushed graphite within a two-week period. In a long-term experiment, cumulative releases were measured periodically from an intact specimen of graphite over a 14 month period. A small fraction of the graphite carbon-14 inventory was released to the gas phase (∼0.004% as CO/CO2 and ∼0.001% associated with organic compounds). A larger quantity of carbon-14, about 0.1%, was released to the solution phase and was thought to be mainly 14CO2, with some possible organic component. In general, the rate of gaseous carbon-14 release decreased with time. The results suggest a small initial release of relatively labile, accessible carbon-14, with longer term release occurring at a much slower rate. Tritium (T) releases were also measured.

The thermodynamic solubility product for baryte, determined at standard conditions, from data in commonly used compilations, was compared with published experimental solubility products for baryte and with the solubility products from databases used by the PHREEQC geochemical speciation code and MultiScale, which is often used by oil companies to predict or describe baryte scaling. The values in the various databases agree well with experimental data (10–10.05–10–9.96; Melcher, 1910; Neuman, 1933; Templeton, 1960; Davis and Collins, 1971; Blount, 1977; Felmy et al., 1990), which agree within uncertainty with the values presented in the compilations of Robie et al. (1979), Wagman et al. (1982), Lide (2005), Raju and Atkinson (1988), as well as Nordstrom and Munoz (1994), whose values have the least uncertainty. In solutions of 50% seawater mixed with 50% reservoir formation waters, the data predict baryte supersaturation, both at standard temperature and at the temperatures expected in the reservoir, completely consistent with field observations. This provides confidence that the Pitzer approach for activity correction and the database is valid for investigations of baryte precipitation rates in high ionic strength solutions.

The implementation of a geological disposal facility requires the demonstration of confidence that such a facility would be safe during both the operational period and in the long-term after the closure of such a facility. The generic environmental safety case described in this paper is the vehicle used to demonstrate an understanding of environmental safety. It will be used to prepare a site-specific environmental safety case in due course. The approach taken will be consistent with a staged development and approval process, as advocated by the environmental regulators.

Atomic force microscopy has been used to determine more precisely the mechanism of the initial stages of dolomite dissolution. Analysis of outflow solutions initially shows fluctuations of both Ca and Mg release with concentrations of Ca >> Mg. The dolomite surface dissolves congruently in the presence of slightly acidified water as confirmed by the regular spreading of characteristic rhombohedral etch pits. Direct in situ observations show that a new phase precipitates on the dissolving surface simultaneously. As the Ca and Mg release decreases with time, the precipitated phase can be seen to spread across the dolomite surface. These observations indicate that the apparent incongruent dissolution of dolomite is a two-step process involving stoichiometric dissolution with the release of Ca, Mg and CO3 ions to solution at the mineral–fluid interface coupled with precipitation of a new Mg-carbonate phase. The coupled element release confirms the interface-coupled dissolutionprecipitation mechanism.

Batch experiments have been used to assess the sorption properties of a potential cementitious repository backfill, NRVB (Nirex reference vault backfill), using Eu3+ as a model trivalent radionuclide and ethylenediaminetetraacetic acid (EDTA) as a competing ligand. The NRVB is an effective scavenger of Eu from solution, with most sorbed within minutes onto the crushed material and less than 1.5% remaining in solution after one day (R d values in the range 0.3–2.4 × 104 l kg−1). Ultrafiltration showed that nearly all of this remaining Eu (>94%) is attached to NRVB derived colloids or particulates that are mainly retained by a 100 kDa ultrafilter. High concentrations of EDTA (>0.01 M) reduced the extent of sorption at apparent equilibrium. The addition of EDTA to a pre-equilibrated system of Eu3+ and NRVB resulted in a temporary suspension of some Eu, but this very quickly returned to the solid phase. There is some irreversibility in these systems, with EDTA able to prevent removal of Eu(III) from solution, but unable to bring it back into solution under the same conditions.

This paper gives an overview of the geosphere research studies being undertaken by the Radioactive Waste Management Directorate (RWMD) of the Nuclear Decommissioning Authority. The approach of the RWMD in the current generic phase of the UK managing radioactive waste safely (MRWS) programme is to maintain an understanding of key processes and to carry out research and development into techniques so capability can be built. Although RWMD can demonstrate a general understanding of geosphere processes at this stage in the UK project, it is recognized that this will need to be made site-specific as the MRWS programme progresses. An understanding of the geosphere at the selected site(s) will be an important part of the future programme. Where possible, the RWMD will participate in international studies so that relevant site-based information can be accessed. In this way, the RWMD will be prepared for site-specific work in stage 5 of the MRWS process.

Samples of artificially aged Nirex reference vault backfill (NRVB) were prepared by progressive leaching with deionized water, after which some of the samples underwent hydrothermal treatment. Compared to unaltered NRVB, X-ray diffraction (XRD) and differential scanning calorimetry (DSC) provided evidence for the ageing resulting in alteration of the mineralogy, in particular the absence of portlandite was observed. The specific surface area of NRVB initially increased due to leaching, but then decreased after further leaching.

Sorption distribution ratios (R D values) of uranium(VI), neptunium(IV), tin and zirconium onto aged NRVB samples were measured using the batch sorption technique. For all four elements, there was little difference between R D values for aged NRVB and those for untreated material. The most probable explanation for these findings is that even though the ageing treatments altered the NRVB mineralogy, calcium silicate hydrate (C-S-H) phases are responsible for most of the radionuclide uptake and their transformation to more crystalline forms has little effect on the bulk sorption capacity of the aged material.

Gases will be generated in waste packages during their transport to a geological disposal facility (GDF), this generation will continue during GDF operations and after GDF closure. The range of gases produced will include flammable, radioactive and chemotoxic species. These must be managed to ensure safety during transport and operations, and the post-closure consequences need to be understood. The two primary post-closure gas issues for a GDF are the need for the system pressure to remain below a value at which irreversible damage to the engineered barrier system and host geology could occur, and the need to ensure that any flux of gas (in particular gaseous radionuclides) to the biosphere does not result in unacceptable risk. This paper provides an overview of the research of the Nuclear Decommissioning Authority, Radioactive Waste Management Directorate into gas generation and its migration from a GDF.

The Cyprus Natural Analogue Project was carried out due to the requirement to support ongoing laboratory and modelling efforts on the potential reaction of the bentonite buffer with cementitious leachates in the repository engineered barrier system. Although it is known that the higher pH (12.5–13) leachates from ordinary Portland cement will degrade bentonite, it is unclear if this will also be the case for the lower pH (10–11) leachates typical of low alkali cements. Ongoing laboratory and underground rock laboratory programmes, which are currently investigating this, face the obstacle of slow kinetics and the production of short-lived metastable phases, meaning obtaining unambiguous results may take decades. It was therefore decided to implement a focussed natural analogue study on bentonite/low alkali cement leachate reactions to provide indications of the probable long-term reaction products and reaction pathways to provide feedback on the existing short-term investigations noted above and to ascertain if any critical path research and development needs to be instigated now. The results of the analyses presented here, in this short overview of the project, suggest that there has been very limited alkaline groundwater reaction with the bentonite. This is generally supported by both the geomorphological evidence and the natural decay series data which imply groundwater/rock interaction in the last 105 a.

The process of transforming hydromagnesite to magnesite is analysed in the context of the theory of solvent-mediated transformations. A series of experiments at 120, 150 and 180ºC with different heating times was designed to determine, by powder X-ray diffraction, the amount of magnesite generated as a function of time. The aqueous-phase composition was monitored by inductively coupled plasma-mass spectrometry and carbonate alkalimetry. From the analytical data, the evolution of saturation indexes with respect to both phases was determined using the geochemical code PHREEQC. Finally, two different methods were applied to obtain the activation energy of the process and a TTT (Temperature- Transformation-Time) graph was constructed to define suitable conditions in which to obtain magnesite.