In the autumn of 2006, the UK government declared its expectation to set up a National Nuclear Laboratory (NNL) as soon as possible (subject to satisfactory contractual terms). The aim would be to base the Laboratory around Nexia Solutions and its ‘state of the art’ facility at Sellafield in Cumbria. The initial phase of the work to recommend formation of the laboratory is well advanced, and the NNL is expected to have a remit for the following roles:

— to play a key role in supporting the UK's strategic R&D requirements

— to operate world-class facilities

— to ensure key skills are safeguarded and enhanced

— to play a key role in the development of the UK's R&D supply base.

The NDA remit as set out within the Energy Act includes – “promote, and where necessary fund, research relevant to nuclear clean up”. The NDA need to underpin delivery and / or accelerate programmes to fulfil the overall mission and technical underpinning of these activities is critical. In this paper we will present consideration of the investment required in nuclear waste Research and Development.

Firstly, NDA set the requirement for nuclear sites to write down within the Life Time Plans (LTP), at a high level, the proposed technical baseline underpinning the LTP activities; furthermore we required technology gaps / opportunities in the technical baselines to be outlined in a R&D requirements section to the LTP. Criteria were established to categorise the R&D in three areas:

“needs” - those development activities needed to underpin the proposed technical solutions

“risks” – those activities required to reduce / eliminate key risks to the proposed technical solutions

“opportunities” – innovations / changes to the technical baselines

The purpose of production of the technical baselines and underpinning R&D requirements is to establish an auditable trail through the LTP from programme components into how the programme will be delivered.

NDA believes the production of the technical baselines and R&D requirements will be of benefit to the Site License Companies (SLC) in terms of ensuring a focus on overall programme delivery and not just short term activities. Furthermore, we can ensure that investment in technology is targeted at priority areas, with common issues and requirements identified and solutions on a broader scale will be achievable.

This presentation gives an overview of the French major research program on nuclear waste and presents some perspectives in the context of the 2006 Act, while focusing on the main CEA contributions. Development of new conditioning processes combined with in-depth studies of the waste-packages long term behaviour studies allows enhancing the expected performance of waste-packages. Geological disposal performances have been demonstrated based in particular on the relevant understanding of the radionuclides chemistry and migration in complex environment, including tracing field-experiments in ANDRA underground research lab. French research and plans for waste from future reactors will be discussed. It is still a very prospective area as fuel materials that will be used in these reactors is not decided. The capacity to recycle most of the fuel component will be a major point in the choice of an integrated concept including reactor and fuel cycle.

The UK's intermediate level waste (ILW) comprises a range of waste types, which vary widely in terms of physical and chemical properties, radionuclide content (including fissile) and half-life. ILW needs to be conditioned in such a way as to provide confidence that a future disposal facility for such wastes will be able to meet or exceed the basic long-term safety and environmental protection standards. The reactive properties and heterogeneous nature of certain ILW can present waste immobilisation challenges and, depending on the waste conditioning process that is adopted, may ultimately preclude the production of truly passive wasteforms. Of considerable interest is the extent of long-term containment likely to be afforded by the waste container, the wasteform and any engineered barriers within the disposal system. A regulatory view of the current uncertainties is provided and areas where further understanding would be beneficial are identified.

An integral part of decommissioning the Nuclear Decommissioning Agency's (NDA) Dounreay site is the management of the solid low level radioactive waste (LLW). The United Kingdom Atomic Energy Authority (UKAEA) has developed and progressed a technical and stakeholder programme that has enabled it to submit a robust Planning Application to Highland Council (HC) for New LLW Disposal Facilities at Dounreay and to submit substantive preliminary safety and environmental cases to the Nuclear Installations Inspectorate (NII) and the Scottish Environment Protection Agency (SEPA). To UKAEA’s knowledge this is the most advanced project in the UK for new LLW disposal facilities. Experience has been gained in progressing the Best Practicable Environmental Option (BPEO) Study, working with regulators in unfamiliar areas, and undertaking groundbreaking stakeholder consultation. Key lessons learnt are that stakeholders should be engaged in dialogue on the project as early as possible, documentation must be high quality and tailored to its audience, and internationally respected and credible consultants must be involved.

Excess plutonium not destined for burning as MOX or in Generation IV reactors is both a long-term waste management problem and a security threat. Immobilisation in mineral and ceramic-based waste forms for interim safe storage and eventual disposal is a widely proposed first step. The safest and most secure form of geological disposal for Pu yet suggested is in very deep boreholes and we propose here that the key to successful combination of these immobilisation and disposal concepts is the encapsulation of the waste form in small cylinders of recrystallized granite. The underlying science is discussed and the results of high pressure and temperature experiments on zircon, depleted UO2 and Ce-doped cubic zirconia enclosed in granitic melts are presented. The outcomes of these experiments demonstrate the viability of the proposed solution and that Pu could be successfully isolated from its environment for many millions of years.

Cubic zirconia is a well known, highly durable material with potential uses as an actinide host phase in ceramic waste forms and inert matrix fuels and in containers for very deep borehole disposal of some highly radioactive wastes. To investigate the behaviour of this material under the conditions of possible use, a cube of ∼ 2.5 mm edge was made from a single crystal of yttriastabilized cubic zirconia doped with 0.3 wt.% CeO2. The cube was enclosed in powdered granite within a gold capsule and a small amount of H2O added before sealing. The sealed capsule was held for 4 months in a cold-seal pressure vessel at a temperature of 780°C and a pressure 150 MPa, simulating both the conditions of a deep borehole disposal involving partial melting of the host rock and the conditions under which the actinide waste form might be encapsulated in granite prior to disposal. At the end of the experiment the quenched, largely glassy, sample was cut into thin slices and studied by optical microscopy, EMPA, SEM and cathodoluminescence methods. The results show that no corrosion of the zirconia crystal or reaction with the granite melt occurred and that no detectable diffusion of elements, including Ce, in or out of the zirconia took place on the timescale of the experiment. Consequently, it appears that cubic zirconia could perform most satisfactorily as both an actinide host waste form for encapsulation in solid granite for very deep disposal and as a container material for deep borehole disposal of highly radioactive wastes (HLW), including spent fuel.

In the context of the present Spanish ‘once-through’ nuclear fuel cycle, the need arises to complete the geological repository reference concept with a spent fuel canister final design. One of the main issues in its design is selecting the inner material to be placed inside the canister, between the steel walls and the spent fuel assemblies. The primary purpose of this material will be to avoid the possibility of a criticality event once the canister walls have been finally breached by corrosion and the spent fuel is flooded with groundwater. That is an important role because the increase in heat generation from such an event would act against spent fuel stability and compromise bentonite barrier functions, negatively affecting overall repository performance. To prevent this possibility a detailed set of requirements for a material to fulfil this role in the repository environment have been devised and presented in this paper. With these requirements in view, eight potentially interesting candidates were selected and evaluated: cast iron or steel, borosilicate glass, spinel, depleted uranium, dehydrated zeolites, haematite, phosphates, and olivine. Among these, the first four materials or material families are found promising for this application. In addition, other relevant non-performance-related aspects of candidate materials, which could help on decision making, are also considered and evaluated.

The acoustic emission (AE) technique was used to detect and characterise the processes associated with generation of stress waves during melting and solidification of granite E93/7 at a pressure of 0.15 GPa. The AE signals recorded as a result of partial melting of the granite at a temperature of 780°C and subsequent solidification during cooling were distinguished from the equipment noise and their parameters used to characterise the AE sources associated with the phase transformations during melting and solidification of the granite. The mechanisms generating AE during granite melting were differentiated by AE signals with their highest peaks in the frequency spectrum at 170 and 268 kHz. The transformation of the liquid into glass during solidification of the partially melted granite generated AE waves in an essentially broad range of frequencies between 100 and 300 kHz. This preliminary work demonstrates the potential of the AE technique for use in applications related to deep borehole disposal of radioactive wastes.

We present a mathematical model for determining the temperature field around radioactive waste containers in very deep geological boreholes. The model is first used to predict the temperature rise for some simple, but well-established cases with known solutions in order to verify the numerical work. The temperature distribution is then determined for two variants of the deep bore hole concept; a low temperature variant and a high temperature variant. The results from these studies are discussed in terms of their utility in establishing deep borehole disposal as a workable concept.

Composite cements where large amounts of blast furnace slag (BFS) replace Portland cement are currently used for immobilisation of ILW. Hydration of BFS is activated by the small amount of OPC present but the amount of reaction that occurs is limited at ambient temperatures. Increasing the temperature increases the hydration of the BFS but large amounts still remain unreacted, leaving a porous matrix where the capillary pores remain filled with a highly alkaline solution. This solution causes corrosion of reactive metals giving rise to expansive reactions and hydrogen release, and it can destroy the structure of zeolites releasing the adsorbed species.

Apart from OPC, BFS hydration can be activated by other compounds such as hydroxides, sulphates, silicates, and calcium aluminate cements. The use of these alternatives gives rise to binders such as ettringite and strätlingite which have a different chemistry where the pore solution has a lower pH. Corrosion of metals does not readily occur in these binders. This may be due to the reduced pH but could also arise from the lack of pore water, as these binders bind more water in their structure so that it is not available for transport of ionic species. This extra water binding also has potential for immobilisation of sludges where high w/s ratios are necessitated by the need to transport the sludge.

This paper will review some of the alternative activators for slag hydration and present experimental results on several systems where slag has been activated with compounds other than OPC.

Cementitious material is the most commonly used encapsulation medium for low and intermediate level radioactive waste. This paper focuses on the aqueous durability of a Materials Testing Reactor (MTR) cementitious wasteform – a possible candidate for the proposed intermediate level waste management facility in Australia. A series of medium term (up to 92 months) durability tests, without leachate replacement, were conducted on samples of this wasteform.

The wasteform was made from cement, ground granulated blast furnace slag and a simulated waste liquor. The compressive strength (39 MPa) was typical of MTR cement wasteforms and well above that required for handling or storage. The wasteform was an inhomogeneous mixture containing calcite, a calcium silicate hydrate phase, hydrotalcite and unreacted slag particles. After leaching for 92 months the crystallinity of the calcium silicate hydrate phase increased.

The majority of the releases of Ca, Si, Al, Sr, S, Na and K was reached within 4 days of leaching, with the maxima ie. the highest concentrations in the leachates, occurring at 3 months for Ca, Al, Sr, S, Na and K, and at 1 month for Si. For the longer leach periods (6 months and 3 months respectively) there was a slight reduction in concentration in the leachates, and these levels were similar to those for the longest period of 92 months, suggesting steady-state conditions prevailing after 3 to 6 months of leaching. The highest releases of matrix elements were for Na (37%), K (40%) and S (16%). Releases for elements such as Ca, Na, Al and Sr were similar in magnitude to those reported by the UKAEA in earlier MTR studies.

After leaching for 92 months there was an alteration layer about 80 ∞m deep where calcium has been depleted. Na, K and Sr showed signs of diffusion towards the outer part of the cement samples.

Thermochemical processing of reactor graphite waste is based on self-sustaining reaction 4Al + 3TiO2 + 3C = 3TiC + 2Al2O3 which chemically binds 14C from the irradiated graphite in the titanium carbide. Thermochemical processing was investigated to analyse the behaviour of rare earth elements (REE), where REE = Y, La, Ce, Nd, Sm, Eu and Gd. Both thermodynamic simulations and laboratory scale experiments were used. The REEs in the irradiated reactor graphite are formed as activation products of impurities and spread over the graphite bricks surfaces as well as arise from fission of nuclear fuel. REEs can be used also to substitute for waste actinides as well as to increase the durability of carbide-corundum ceramics relative to waste actinides.

Thermodynamic calculations and X-ray diffraction analysis of ceramic specimens synthesized revealed that durable REE's aluminates with perovskite, β-alumina and garnet structures are formed by interaction of REE oxides with the Al2O3 melt during the selfpropagating reaction of ceramic formation.

The porous carbide-corundum ceramics synthesized have a high hydrolytic durability, e.g. the normalised leaching rates of 137Cs, 90Sr and Nd are of the order of 10–7 – 10–8 g/(cm2·day).

Water is a necessary component in the production of encapsulated wastes based on hydraulic cements which are widely used for immobilization of intermediate and low level waste (ILW) and (LLW). Apart from providing the fluidity required to readily transport slurry wastes, it plays an essential role in hydrating the cement. Too low a water content prevents homogeneous mixing of the cement binder and waste and does not provide the fluidity needed for effective infilling of solid wastes. The water left after hydration creates a porous network that allows egress of gaseous corrosion/radiolytic degradation products such as hydrogen. A broad envelope (i.e. range) of acceptable water/binder ratios is essential for effective process control, particularly for the encapsulation of slurry wastes which have widely varying water contents.

Nevertheless, the presence of large amounts of free water in the pore system of the hardened matrix allows easy transport of soluble ions such as hydroxide, which can lead to metal corrosion, and the increased permeability of the system increases the leachability. Therefore effective management of the ‘free’ water content of a waste product will allow optimisation of both the encapsulation process and the product quality and durability.

This paper describes a range of innovative approaches to ‘water management’, including the use of alternative hydraulic cements, modification of powder characteristics and use of superplasticised composite OPC grouts and examines the contribution of 1H NMR relaxometry in providing improved understanding of the distribution of water within the pores of the hardened cement matrix.

The data on historic mortars and concretes provide qualitative and quantitative information to evaluate long-term behavior of cement materials in repositories and to understand processes that may occur in repositories (e.g., interaction with other materials and radionuclide transfer). Beyond that, such information is important to demonstrate safety aspects of the repositories to the public and stakeholders.

A number of reports have been devoted to study of historical mortars and concretes used in the Western countries. The purpose of this paper is to review studies on compositions and structures of analogs, located mainly over the former Soviet Union's territory.

The zeolitic ion exchanger clinoptilolite was encapsulated within various cementitious systems in order to assess their suitability for the retention of the radioelements, Cs and Sr. The pozzolanic reaction of clinoptilolite is reduced in composites containing BFS and PFA and appears not to continue after 7 days of hydration. Ca(OH)2 persists up to 360 days of hydration in a 9:1BFS:OPC system with 10% clinoptilolite added, despite the presence of unreacted pozzolan. This may be due to low pH of the pore solution, if Na and K act as counter cations in the aluminous C-S-H, a product of pozzolanic hydration or are exchanged onto the clinoptilolite. Saturation of the pore solution with Ca may prevent further dissolution of Ca(OH)2. Cs leaching occurs in all samples during accelerated tests due to breakdown of the clinoptilolite structure.

The alternative cement system calcium sulfoaluminate cement (CSA) has a different hydration chemistry and properties to OPC and OPC composites with a lower pore solution pH. Clinoptilolite appears to react in a hydrating CSA system with significant reaction continuing between 28 and 90 days of hydration. Leaching of Cs from CSA is higher than from an OPC system, in which almost all of the clinoptilolite crystallinity is lost. The major product of CSA hydration is ettringite. Cs may be adsorbed within cation sites of the C-S-H in an OPC system but not by ettringite which does not retain Cs so Cs has high mobility and leachability through the CSA matrix.

The reaction of CO2 gas with OPC, OPC-BFS and OPC-PFA composite cement systems were studied using XRD, SEM and TG to investigate the applicability of these materials to immobilise carbon arising from graphite waste. XRD results suggested that calcite formed in OPC system after the carbonation reaction, whereas calcite and vaterite were observed in OPCBFS and OPC-PFA systems. In OPC system, nearly half of Ca(OH)2 was consumed to form CaCO3. In OPC-BFS and OPC-PFA systems, the amount of CaCO3 formed, corresponded to the consumption of greater than 100% of Ca(OH)2 initially present, suggesting that other hydration products e.g. C-S-H were also consumed, either directly or indirectly during the carbonation process. The OPC-BFS system became more porous after carbonation. OPC-PFA system indicated a high efficiency on the conversion of Ca in the system into CaCO3.

According to the present Belgian radioactive waste management program, Eurobitum bituminised radioactive waste will be disposed of in a geologically stable underground clay formation. The Boom Clay is studied as a potential host formation because of its low diffusion and high retention properties towards radionuclides. The presence of the radioactive waste should not disturb these properties. Due to the presence of hygroscopic salts (25 to 30 weight% NaNO3), Eurobitum will take up pore water which will result in a swelling and possibly in a very high swelling pressure. First scoping calculations suggest that the swelling pressure exerted to Boom Clay should remain below 7 to 8 MPa to avoid the formation of fractures. If the bitumen in EUROBITUM behaved like a perfect semi-permeable membrane and if no swelling were allowed after the dissolution of NaNO3 into a saturated solution of 10.8 M, osmotic pressures of ∼50 MPa could be attained. To better understand the interaction between the swelling Eurobitum and the host formation, coupled hydro-chemical-mechanical constitutive laws for Eurobitum have to be developed. To this purpose, water uptake tests under constant volume (‘confined’) and constant stress (‘semi-confined’) conditions are being performed. After ∼2 years of hydration of small inactive Eurobitum samples in constant volume conditions, the swelling pressure has raised to ∼12 MPa. The volume of samples that can swell against counter pressures of 2.2, 3.3, or 4.4 MPa (constant stress tests) increased with ∼5 to 11 volume%, independently of the applied counter pressure. Approximately 10 weight% of the initial NaNO3 content has been leached.

The decommissioning of the UK’s Magnox reactor sites will produce large volumes of low level waste (LLW) arisings. The vast majority of this waste takes the form of concrete, building rubble and redundant plant containing relatively low levels of radioactivity. Magnox Electric Ltd (Magnox) is leading a strategic initiative funded by the Nuclear Decommissioning Authority (NDA) to explore opportunities for the disposal of such waste to suitably engineered facilities that might be located on or adjacent to the site of waste arising, if appropriate and subject to regulatory acceptance and stakeholder views. The strategic issues surrounding this initiative are described along with an update of progress with stakeholder consultations in relation to the proposed licensing of the first such facility at Hinkley Point A, which could be viewed as a test case for the development of similar disposal facilities at other nuclear sites in England and Wales.

The operation of nuclear fuel reprocessing plants generates a radioactive effluent containing nitrate salts as the major constituent. This waste must be disposed of safely and economically, and to achieve this aim the Japan Atomic Energy Agency (JAEA) is developing a cement based encapsulation method to immobilise this waste. Non-radioactive development work has been performed at both small and large-scale (up to 200 litres) to investigate the optimum cement formulation. The results from these studies demonstrate that nitrate waste that has been concentrated by evaporation of the water to a predetermined level can be successfully encapsulated up to a waste loading corresponding to 50 wt% sodium nitrate. It has been identified that high concentrations of bicarbonate ions in the effluent can have a detrimental effect on the strength of the cement encapsulated waste; however, provided this concentration is controlled, successful encapsulation of the effluent is still achieved.