A 65 m vertical shaft was sunk at Dounreay in the 1950s to build a tunnel for the offshore discharge of radioactive effluent from the various nuclear facilities then under construction. In 1959, the Shaft was licensed as a disposal facility for radioactive wastes and was routinely used for the disposal of ILW until 1970. Despite the operation of a hydraulic containment scheme, some radioactivity is known to have leaked into the surrounding rocks. Detailed logging, together with mineralogical and radiochemical analysis of drillcore has revealed four distinct bedding-parallel zones of contamination. The data show that Sr-90 dominates the bulk beta/gamma contamination signal, whereas Cs-137 and Pu-248/249 are found only to be weakly mobile, leading to very low activities and distinct clustering around the Shaft. The data also suggest that all uranium seen in the geosphere is natural in origin. At the smaller scale, contamination adjacent to fracture surfaces is present within a zone of enhanced porosity created by the dissolution of carbonate cements from the Caithness flagstones during long-term rockwater interactions. Quantitative modelling of radionuclide migration, using the multiphysics computer code QPAC shows the importance of different sorption mechanisms and different mineralogical substrates in the Caithnesss flagstones in controlling radionuclide migration.

Some crystals doped with radionuclides glow in the dark. Such materials are prospective for certain industrial scale applications. Durable self-glowing crystalline solids, which were initially suggested for development of actinide waste forms, are considered as advanced materials. Well-known durable actinide host phases, such as zircon, xenotime, and monazite are main focus of current research. Single crystal samples of these host phases doped with 239Pu, 238Pu, 241Am and 237Np have been grown by flux methods. It is demonstrated that incorporation of small amounts of non-radioactive elements such as Eu, In and Tb increases the self-glowing intensity. The optimal content of such luminescence ions supporting intensive glowing of 238Pu-doped zircon and xenotime has, at first, been identified by cathodoluminescence study of non-radioactive samples. Subsequently, the results of this study were used to grow intensively glowing crystals of zircon and xenotime doped with 0.01 wt. % and 0.1 wt.% 238Pu, respectively.

Coffinite (USiO4, I41/amd, Z=4) is the major alteration phase of uraninite, UO2+x, under reducing conditions in natural uranium deposits. Thus, it is important to understand the radiation response of coffinite because it is an expected alteration product of the UO2 in spent nuclear fuel. In the present study, we conducted in-situ transmission electron microscopy (TEM) investigation of synthetic coffinite under 1 MeV-Kr2+ ion beam irradiation. The radiation-induced crystalline-to-amorphous transformation was observed in the synthetic nanocrystalline USiO4, with a critical dose of ∼ 0.27 displacements per atoms (dpa) for which full amorphization occurred at room temperature. The critical dose increases as rising irradiation temperature, and above the critical temperature (Tc), ∼ 608 K, coffinite cannot be amorphized. These results are compared with previous studies on the isostructural zircon (ZrSiO4, Tc=1000K) and thorite (ThSiO4, Tc>1100K), which indicates that synthetic coffinite is more stable to ion beam irradiation at elevated temperature.

Single crystals and polycrystalline samples of durable actinide host phases such as garnet, zircon, and Ti-pyrochlore doped with varying amounts of U, Pu, Np and Am, were studied by cathodoluminescence (CL) spectroscopy. The following characteristic bands were identified: uranyl-ion [UO2]2+– 495 nm in garnet and pyrochlore; [UO4]2− – 710 nm in garnet; and Am3+ – in garnet and zircon at 550 nm, 614 nm, 740÷750 nm.

South Africa initiated a program to significantly expand its total nuclear energy capacity by acquiring additional Pressurized Water Reactors (PWRs) and is currently researching Pebble Bed Modular Reactors (PBMRs) as next generation nuclear energy sources. The result of such an expansion in the current nuclear program will be a need for the disposal of additional nuclear waste. However, as an alternative strategy for spent PBMR fuel, the separation of the spent fuel into TRISO particles for disposal and graphite for re-use in the nuclear industry could also be considered.

Immobilization of long-lived 99Tc requires development of chemically resistant inorganic matrices. Samples of ceramics based on crystalline Fe-Mn- and Zr-Mn-oxide compounds were synthesized at 1150°C in air, reducing or inert atmosphere from precursors doped with 5-12 wt.% Tc. All the samples obtained were studied using optical and scanning electron microscopy (SEM); powder X-ray diffraction (XRD) and microprobe analysis (EMPA). Content of Tc varied from 0.5-0.8 to 3-6 wt.% in oxide host phases and from 54 to 93 wt.% in metallic inclusions. It was demonstrated that synthesis of oxide host-phases under oxidizing or reducing conditions was not optimal due to partial Tc volatilization or metallic phase formation, respectively. The use of inert atmosphere for ceramic synthesis supports Tc incorporation into crystalline structure of stable host-phases. Development of optimal methods of precursor preparation and synthesis conditions of Tc-doped ceramic are being discussed.

Experiments on determination of operational effectiveness of track membranes with different diameter of pores (0.2 – 0.05 microns) for additional purification of water of the first contour of the reactor before ion-exchange resins were carried out. The track membranes detain products of corrosion, and also products of wearing of moving parts of the equipment by the size of > 0.05 microns. Filtration experiments on of water of the first contour of the exploratory reactor of VVR-SM have shown high performance of track membranes.

The technology of removing cesium radionuclides from sodium coolant at the BN-350 fast reactor was realized in the form of two different types of cesium traps: stationary devices connected to the circuit and in-core devices installed into the core of reactor when it was not under operation. Carbon-graphite materials were used as sorbents in these traps to collect and concentrate radioactive cesium, accumulated in the BN-350 reactor circuits over the decades of their operation. The relatively small volume traps provided effective radiation-safe conditions for personnel working close to the primary circuit coolant and equipment during BN-350 decommissioning. These spent cesium traps represent solid radioactive wastes that must be treated before long term storage. The presence of chemically active sodium, potassium and cesium in the traps results in series of problems related to their long term storage and disposal in the Republic of Kazakhstan. As a consequence, the technology of filling spent cesium traps with lead/lead-bismuth alloy was evaluated. A set of experiments was implemented aimed at verification of calculations performed in substantiation of the proposed technology: filling a full scale cesium trap mock-up with sodium followed by its draining to determine the optimal regimes of draining; filling bench scale cesium trap mock-ups with sodium and cesium followed by sodium draining and filling with lead or lead-bismuth alloy at different temperatures and filling rates to chose the optimal regimes for filling spent cesium traps; implementation of leachability tests to determine the rate of cesium release from the filling materials into water. This paper provides a description of the experimental program carried out and the main results obtained.

Radiation damage and the effect on physical and chemical properties is an important component in the prediction of the long-term stability of waste form materials. As part of the ongoing goal of increasing the accuracy of long-term predictions of radiation damage, two types of material, based on proposed materials with a waste form application have been irradiated. Results have shown that Y2TiO5 (Y2.67Ti1.33O6.67), and Yb2TiO5 (Yb2.67Ti1.33O6.67), both of which are non-stoichiometric, disordered pyrochlore-based compounds, behave significantly different to the stoichiometric, ordered pyrochlore equivalent. For example the critical temperature, the temperature above which materials remain crystalline during irradiation, is found to decrease from the ordered equivalents, e.g. Y2Ti2O7. A second material based on La2TiO5 has been found to behave differently to both La2/3TiO3 and La2Ti2O7, with a change in Tc of ∼ 200 K.

The focus of this research programme is to develop a single phase ceramic wasteform for waste PuO2 that is unsuitable for fuel manufacture. A suite 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 of each system.

An understanding of the effect of cumulative radiation damage on the integrity of ceramic wasteforms for plutonium and minor actinide disposition is key to the scientific case for safe disposal. Alpha recoil due to the decay of actinide species leads to the amorphisation of the initially crystalline host matrix, with potentially deleterious consequences such as macroscopic volume swelling and reduced resistance to aqueous dissolution. For the purpose of laboratory studies the effect of radiation damage can be simulated by various accelerated methodologies. The incorporation of short-lived actinide isotopes accurately reproduces damage arising from both alpha-particle and the heavy recoil nucleus, but requires access to specialist facilities. In contrast, fast ion implantation of inactive model ceramics effectively simulates the heavy recoil nucleus, leading to amorphisation of the host crystal lattice over very short time-scales. Although the resulting materials are easily handled, quantitative analysis of the resulting damaged surface layer has proved challenging.

In this investigation, we have developed an experimental methodology for characterisation of radiation damaged structures in candidate ceramics for actinide disposition. Our approach involves implantation of bulk ceramic samples with 2 MeV Kr+ ions, to simulate heavy atom recoil; combined with grazing incidence X-ray absorption spectroscopy (GI-XAS) to characterise only the damaged surface layer. Here we present experimental GI-XAS data acquired at the Ti and Zr K-edges of ion implanted zirconolite, as a function of grazing angle, demonstrating that this technique can be successfully applied to characterise only the amorphised surface layer. Comparison of our findings with data from metamict natural analogues provide evidence that heavy ion implantation reproduces the amorphous structure arising from naturally accumulated radiation damage.

Two different methods have been used to synthesize sodalite for conditioning of chloride salt wastes coming from pyroprocesses: the first one, starting from kaolinite through the intermediate nepheline phase; the second one, starting from silica and sodium aluminate reagents, directly. The obtained products have been characterized by means of several analyses. In particular, different instrumental techniques – stereomicroscopy, scanning electron microscopy (SEMEDS), density measurements, thermogravimetric analysis, X-rays diffraction, FTIR spectroscopy – were performed revealing that the synthesis from kaolinite is the best method, provided that rigorous conditions are followed. The use of an argon atmosphere for the preparation of pellets of reagents is strictly necessary for the obtainment of a good quality product.

The choice of surrogate for plutonium oxide for use during the initial stages of research into the immobilization of intermediate level pyrochemical wastes containing plutonium andamericium oxides in a calcium phosphate host has been investigated by powder X-ray diffraction and X-ray absorption spectroscopy. Two non-radioactive surrogates, hafnium oxide and cerium oxide, together with radioactive thorium oxide were compared. Similarities in behaviour were observed for all three surrogates when calcined at the lowest temperature, 750°C but differences became more pronounced as the calcination temperature was increased to 950°C. Although some reaction occurred between all the surrogates and the host to form a substituted whitlockite phase, increasing the temperature led to a significant increase in the cerium reaction and the formation of an additional phase, monazite. Additionally it was observed that the cerium became increasingly trivalent at higher temperatures.

The paper describes the studies of the transformation of Cs+- and Sr2+-containing zeolite sorbents synthesized from fly ash cenospheres to crystalline mineral composition, suitable for the long-term disposal. Series of Cs+- and Sr2+-exchanged NaP1-containing sorbents were subjected to the thermochemical transformation in the temperature range 40-1100°C at atmospheric pressure in air and the progress of reaction was monitored by DSC and XRD analysis. It was shown that initial sodium zeolite undergoes two-step transformation at 736-785°C and 892-982°C forming nepheline as the principle product, with the conversion temperatures being dependant on the heating rate.

The thermal treatment of Cs+-bearing zeolite sorbent led to formation of a complex multiphase system, the principal components of which were nepheline and pollucite. Increasing cesium content in the samples led to a monotonous shift of crystallization peak to the higher temperature range (1005-1006°C). A more complicated behavior was observed for Sr2+-containing samples, for which the crystallization temperature tends to increase (compared with NaP1) at lower Sr contents, but it starts decreasing parallel to the Sr2+ content at Sr2+ loadings >10 mg/g. The principal crystalline phases in Sr-NaP1 sample conversion were nepheline and Sr2+-containing feldspar, the quantity of which increased parallel to the increase of strontium content in zeolite.

Apparent activation energies of thermochemical transformations were calculated and possible approaches to reduce transformation temperature are discussed and experimentally illustrated.

Fluorapatite (Fap), Ca10(PO4)6F2, could be a potential candidate for the immobilization of radioactive isotopes thanks both to its physical-chemical properties and to the capability to incorporate mono-, di- and trivalent cations in its crystal structure. In this research work different preparation procedures able to incorporate in a fluorapatite matrix strontium in the form of nitrate or fluoride salt were considered. The experiments were performed by means of stable isotopes, as representative of the radionuclides contained in the nuclear fuel. Subsequently, the Sr-substituted fluorapatites (SrxFap), Ca10-xSrx(PO4)6F2, were characterized by X-ray diffraction, Raman spectroscopy, Scanning Electron Microscopy (SEM) and Energy Dispersive Spectroscopy (EDS). The results suggest that the considered synthetic procedures could be reasonably applied to the confinement of the radionuclides contained in some types of nuclear waste. In order to avoid the presence of unreacted reagents in the products or the formation of undesired phases, an optimization of the synthetic procedures may be considered.

In Belgium, Eurobitum intermediate-level long-lived bituminized radioactive waste containing large amounts of NaNO3, which is a hygroscopic and soluble salt, is to be disposed of in an underground repository in a geologically stable clay formation. The Boom Clay is studied as a potential host formation because of its favourable properties to limit and delay the migration of the leached radionuclides and other contaminants (heavy metals, NaNO3, organic molecules) to the biosphere. The emplacement of the bituminized waste will induce multiple processes that could have a significant effect on the key properties of the clay. Because several of these processes are interdependent, the study of the compatibility of Eurobitum with geological disposal is complex. To structure the research and to identify possible knowledge gaps, the Belgian Radioactive Waste Management Agency ONDRAF/NIRAS developed a new methodology based on safety functions and safety statements. In this paper, this methodology is briefly explained, with reference to the disposal of Eurobitum. Experimental results obtained at the Belgian Nuclear Research Centre SCK•CEN are presented and discussed in the light of the safety functions and safety statements approach. The importance of the interdependence of the processes is highlighted. Special attention is given to the evolution of the disposal design as a result of the improved understanding of key processes.

In recent years the increasing oil prices and the need for carbon-free energy to limit global warming have resulted in a revival of interests in nuclear energy. Advanced nuclear fuel cycles are being studied worldwide. They aim at making more efficient use of the available resources, reducing the risk of proliferation of nuclear weapons, and facilitating the management of the resulting radioactive waste. Recently, the Red-Impact project has investigated the impact of a number of representative advanced fuel cycles on radioactive waste management, and more specific on geological disposal. The thermal output of the high-level waste arising from advanced fuel cycles in which all the actinides are recycled is reduced with a factor 3 for a 50 years cooling time and with a factor 5 for a 100 years cooling time in comparison with the spent fuel arising from the once-through fuel cycle. This reduction of the thermal output allows for a significant reduction of the length of the disposal galleries and of the size of the repository. Separation of Cs and Sr drastically reduces further the thermal output of the high-level waste, but it requires a long-term management of those heat generating separated waste streams, which contain the very long-lived 135Cs. Recycling all the actinides strongly reduces the radiotoxicity in the waste, resulting in significantly lower doses to an intruder in the case of a human intrusion into the repository. However, the reduction of radiotoxicity has little impact on the main safety indicator of a geological repository, i.e. the effective dose in the case of the expected evolution scenario; for disposal in clay formations, this dose is essentially due to mobile fission and activation products. The deployment of advanced fuel cycles will necessitate the development of low activation materials for the new nuclear facilities and fuels and of specific waste matrices to condition the high-level and medium-level waste streams that will arise from the advanced reprocessing plants.

Most designs of disposal facilities for higher toxicity radioactive wastes were developed for generic feasibility assessment and hence tend to be rather simple in terms of layout and conservative in the choice of engineered barrier materials. Recently, there has been a trend to reassess such designs in the light of moves towards implementation, where robustness in terms of post closure safety needs to be balanced against the requirements to assure safety, quality, and economic practicality during operation, to minimise environmental impact and to gain public acceptance. Studies to date have, however, tended to focus predominantly on variants of layout and engineered barrier geometry. Evaluation of alternative materials has received less consideration, despite the huge advances in materials science over the last couple of decades and likely further developments in the time until repositories become operational. This paper will examine the constraints that led to the choice of the most common materials selected for barriers within “wet” host rocks and examine the extent to which performance could be optimised by use of alternative materials.

The development of the wireless transmission monitoring in the geological disposal repository, the underground laboratories, the test facilities and the demonstration area, will present an opportunity to provide greater flexibility and robustness to the confirmation of the barrier performance and therefore safety of repository [1]. So as to realize the wireless transmission monitoring especially in the betonite barrier, RWMC has been researching and developing the wireless data transmission system which uses the very low frequency (VLF: around 1 kHz) electromagnetic wave. In this system, transmitting and receiving is conducted through loop coil antennas. The overall transmission system is digitized and maximum 19 sensors can be connected to one transmitter. Transmission distance can be selected from several meters to several hundred meters.

This paper firstly presents the behavior of VLF electromagnetic wave propagation in the rock mass and the underground facility revealed by the in-situ transmission experiment and the analytical method evaluating the behavior of the VLF electromagnetic field in various environments. Development results of the transmitting antenna and the improvement of the transmission performance are secondly shown. The technical feasibility and the future R&D plan of the wireless monitoring in the repository are also explained.

The Centre of Experimental Geotechnics, CTU in Prague has recently launched a sprayed backfill technology development programme. Such technology will, most probably, be used during the future construction of deep underground radioactive waste repositories.

The safe disposal of radioactive waste in deep underground geological repositories assumes the construction of a multi-barrier system the aim of which will be to eliminate the potential penetration of hazardous radionuclides into the biosphere over a time span of several hundred thousand years.

With regard to the location in the repository in which the containers will be placed, bentonite will seal the space between the container and the rock mass in the form of a highly compacted bentonite block, while access galleries to the nests will be backfilled with a mixture of natural bentonite and rock. This mixture will be compacted in layers using compacting machinery thus ensuring maximum backfill density during compaction. Following compaction, however, spaces will remain in the gallery vault to which such compacting machines will be unable to gain access. Sprayed clay (sprayed backfill) technology, however, will allow the filling of such voids by injecting backfill so that the material compaction level obtained using this method is at least the same as the backfill compaction level resulting from that using standard compacting machinery (vibration roller, vibration plate compactor etc.).

Sprayed backfill technology is based on shotcrete technology which is a method commonly applied today in the construction of underground structures. Sprayed backfill technology development may involve the use of concrete spraying machines; however the parameter requirements for the spraying of bentonite backfill differ in principal from those which apply to shotcrete.

The article will provide information on the development of sprayed backfill technology, the machinery employed and the sprayed material as well as research results obtained to date.