The clean-up of the Fukushima Daiichi site, after the March 2011 earthquake and tsunami, continues to generate large amounts of spent adsorbents. These adsorbents need to be disposed of permanently in a low temperature immobilisation process to avoid volatilising radioactive Cs and Sr species. To this end an immobilisation process with a maximum temperature of 600 °C was developed by sintering model waste with glass frit to form a dense Glass Composite Material (GCM) wasteform. A zeolitic model wasteform, chabazite, was sintered with a lead borosilicate glass composition at a maximum temperature of 600 °C. The sintering process was optimised with various thermal treatment steps to ensure that volatile species, aqueous or otherwise, were released before full sintering to yield a dense final wasteform. With this process dense wasteforms of up to 40 wt. % chabazite have been achieved.

Magnesium alloys (Mg-0.5%Zr and Mg-1.2%Mn) and graphite from spent nuclear fuel that has been used in the former French gas cooled reactors, have been stored together in AREVA La Hague plant. The recovery and packaging of these wastes is currently studied and several solutions are under consideration. One of the developed solutions would be to mix these wastes in a grout composed of industrially available cement, e.g. OPC, OPC blended with blast furnace slag or aluminous cement. Within the alkaline pore solution of these matrixes, magnesium alloys are imperfectly protected by a layer of magnesium hydroxide (Mg(OH)2, Brucite) resulting in a slow process of corrosion releasing hydrogen. As the production of this gas must be considered for the storage safety, and the quality of wasteform, it is important to select a cement matrix capable of lowering the corrosion kinetics of magnesium alloys. This is especially true when magnesium alloys are conditioned together with graphite wastes. Indeed, galvanic coupling phenomena may increase early age corrosion of the mixed wastes, as magnesium and graphite will be found in electrical contact in the same electrolyte. Many types of cements have been tested and most of them have caused strong hydrogen production when magnesium alloys and graphite are conditioned together into such cement pastes. Exceptions are geopolymer binder which is already known for that and another binder based on alkali hydroxide activated ground granulated blast furnace slag (AHABFS) which is presented in the present article. First are presented hydrogen production experiments that demonstrate the efficiency of AHABFS towards reduction of corrosion of Mg alloys embedded. In a second part, a formulation of fluid mortar based on this binder is proposed.

In Belgium, Eurobitum bituminized radioactive waste is an important intermediate-level long-lived waste form that contains, besides bitumen and radionuclides, large amounts of soluble salts (mainly NaNO3). Geological disposal in a water-saturated sedimentary formation will induce swelling of Eurobitum due to water uptake by the hygroscopic salts embedded in a highly efficient semi-permeable bitumen membrane. Initially, while there is still free space in the primary waste containers, free swelling will occur. To improve our understanding of the water uptake processes under free swelling conditions, water uptake and leaching tests are being performed in which inactive Eurobitum was contacted at the top with 0.1 M KOH (simplified representation of young cement water) without effective stress on the sample. Under these test conditions, the evolution of the water uptake and NaNO3 leaching processes shows several stages: (1) initial fast ingress of water into and leaching of NaNO3 from pores close to the surface; (2) slower progression of the hydration front, the swelling and the leaching of NaNO3, but behaving linearly with the square root of time, indicating that these processes are controlled by diffusion; (3) after more than 2 to 3 years, the leaching, hydration and swelling rates re-increase, attributed to the formation of interconnecting pores and (micro)cracks in the bitumen matrix in response to its increasing deformation.

In Belgium, the preferred long-term management option for Eurobitum bituminized ILW is its final disposal in a geologically stable clay formation such as the Boom Clay, which is studied as a reference host formation. After disposal, clay pore water will infiltrate the secondary concrete waste containers filled each with ten Eurobitum drums. Eurobitum contains hygroscopic salts, mostly NaNO3 (20-30 wt%) and CaSO4 (4-6 wt%), and thus will take up water and swell. If swelling is hindered, a pressure will be exerted on the concrete container and ultimately on the surrounding Boom Clay, possibly inducing stresses in the clay close to the disposal galleries. To improve our understanding of these processes, water uptake tests are ongoing in which inactive Eurobitum is contacted with 0.1 M KOH (representing young cement water). These tests suggest that the swelling is mainly driven by osmosis. This understanding was validated in the presented research by varying the water activity of the leachant in water uptake tests in both constant stress and constant volume conditions. After a stable swelling rate was reached in contact with 0.1 M KOH, the leachant was switched in the following order: nearly saturated (∼7.8 M) NaNO3 – 0.1 M KOH – nearly saturated NaNO3 – 4 M NaNO3 – 0.1 M KOH. The changes in swelling rate and pressure evolution correlated nicely to the changes in water activity. This confirms that osmosis is the key process governing the swelling of Eurobitum.

Dismantling of old reactors and the management of radioactive graphite wastes are becoming increasingly important issues for a number of IAEA Member States. Exchange of information and research cooperation in resolving identical problems between different institutions contributes towards improving waste-management practices, their efficiency, and general safety. The IAEA Coordinated Research Project (CRP) under the title ’Treatment of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal’ was conducted during 2010-2014 and has involved 24 organisations from ten Member States [1]. The CRP has explored both innovative and conventional methods for graphite characterisation, retrieval, treatment, and conditioning technologies and produced an IAEA technical document [2] which has identified a number of unresolved scientific and technical issues such as the need to:

• 1. Improve the scientific understanding required on creation, chemical form, location and release behaviour (transport models) of radionuclides;

• 2. Improve predictive models of radioisotope behaviour;

• 3. Ensure that sampling programmes are statistically representative of the totality of the graphite to be disposed of;

• 4. Establish an accurate radionuclide inventory;

• 5. Consider novel alternative dismantling and treatment strategies.

The CRP promoted the exchange of technical information on R & D activities and will facilitate practical application for treatment and conditioning of graphite waste. The collaboration continues under the IAEA International Decommissioning and Predisposal Networks (IDN and IPN).

Large stores of unstable waste uranic materials such as fluorides or nitrates exist internationally due to legacy civil nuclear enrichment activities. Conversion of these uranic materials to layered metal uranates prior to disposal is possible via aqueous quench - precipitation type reactions. Previous studies1 have shown facile in-situ formation of geologically persistent and labile uranate colloids2 under simulated nuclear waste repository conditions, though the effects of local solution metal-uranium ratios on uranate stoichiometry have yet to be covered. This affects our understanding of how key radionuclides present in repository porewaters such as strontium or caesium may be sequestered in these uranate structures. In this work, we demonstrate a synthesis reaction for calcium monouranate particles via rapid anhydrous curing of a sol-gel. We present some results showing aqueous nucleation of uranate nanoparticles and their phase transformations during thermal curing as well as the effects of solution phase calcium loading on uranate phase purity in the cured particles.

In the frame of minor actinide recycling, (U,Am)O2 are promising transmutation targets. To assess the thermodynamic properties of the U-Am-O system, it is essential to have a thorough knowledge of the binary phase diagrams, which is difficult due to the lack of thermodynamic data on the Am-O system. Nevertheless, an Am-O phase diagram modelling has been recently proposed by Gotcu. Here, we show a recent investigation of the Am-O system using in-situ High Temperature X-ray Diffraction under controlled atmosphere. By coupling our experimental results with the thermodynamic calculations based on the Gotcu model, we propose for the first time a relation between the lattice parameter and the departure from stoichiometry.

In the framework of the Euratom FIRST-Nuclides program, SCK•CEN conducted leaching experiments on a PWR UOX fuel, taken from the Belgian Tihange 1 reactor, with an average rod burnup of 50 MWd/kgHM in order to determine the rapid release of some of the most critical radionuclides. Experiments were conducted with an intact and a declad spent fuel segment in contact with a reference carbonate solution. During one year, samplings were performed to determine the release of the most relevant radionuclides. The samples taken after six months of leaching were analysed by means of Thermal Ionisation Mass Spectrometry (TIMS). These results were compared to the total inventory measured with radiochemical burnup analyses, as well as with the calculated inventory profile along the pellet radius. The comparison gives information about the origin of the dissolved uranium.

Helium is produced in spent nuclear fuel by α-decays of actinides. After 10,000 years, the concentration of He accumulated in UO2 spent fuel is about 0.23 at.%. For direct disposal of spent nuclear fuel, consequences of helium build-up on the fuel matrix microstructure must be evaluated since it can modify the radionuclide release when water comes into contact with the spent fuel surface, after breaching of the disposal canister. An operational model has been proposed in order to evaluate the effect of helium on the microstructure of spent fuel in a repository. Based on conservative assumptions and different scenarios of bubble population, the calculated helium critical concentration, that could lead to a partial loss of integrity of the spent fuel pellet, is 0.37 at.%. However, observations on He-implanted UO2, α-doped UO2 pellets and natural analogues evidence a macroscopic damage only for He concentrations, which are more than one order of magnitude higher.

Raman spectroscopy studies have been performed on one hand to identify different materials related to spent nuclear fuel (SNF), and on the other hand to study the behavior of SNF at different storage conditions. Specifically, the expected oxidation of the SNF matrix under dry storage conditions and the formation of secondary phases (SP), as a result of corrosion of SNF in a deep geological repository, have been studied. In order to perform these experiments, two protocols based on the Raman spectroscopy technique have been developed. The results show U4O9/U3O7 and U3O8 as oxidation products of UO2 powder at high temperatures in air, and the secondary phase formation (rutherfordine, UO2(CO3), soddyite, (UO2)2SiO4•2H2O, uranophane alpha Ca(UO2)2(SiO3OH)2•5H2O and kasolite, PbUO2SiO4•H2O), due to uraninite corrosion under the conditions of Sierra Albarrana (Spain).

The long-term behavior of the UO2 fuel matrix under conditions of the Belgian “Supercontainer design” was investigated by dissolution tests of high burn-up spent nuclear fuel (SNF) in high alkaline solution under 40 bar of (Ar + 8%H2) atmosphere. Four fragments of SNF, obtained from a pellet previously leached during two years, were exposed to young cement water with Ca (YCWCa) under 3.2 bar H2 partial pressure in four single/independent autoclave experiments for a period of 59, 182, 252 and 341 days, respectively. After a decrease of the concentration of dissolved 238U, which is associated with a reduction of U(VI) to U(IV), the concentration of 238U in solution is constant in the experiments running for 252 and 341 days. These observations indicate an inhibition of the matrix dissolution due to the presence of H2. A slight increase in the concentration of 90Sr and 137Cs in the aqueous solution indicates that there is still dissolution of the grain boundaries. These findings are similar to those reported for spent nuclear fuel corrosion in synthetic near neutral pH solutions.

A trend in the operation of nuclear power reactors is the increase in discharge burnup of the fuel. Intrusion of groundwater in a failed canister of a future deep repository is expected to occur in the presence of hydrogen, produced by the anoxic corrosion of iron and by radiolysis of water. Compelling evidence now exists that hydrogen inhibits oxidative dissolution of irradiated nuclear fuel or alpha doped UO2, however the number of studies involving high burnup fuel are limited. In the experiment PWR UO2 fuel with a burnup of ∼75 MWd/kgU was leached in simulated groundwater (10 mM NaCl, 2 mM NaHCO3) at room temperature under 5 MPa of hydrogen. Since the main objective was to investigate the fuel matrix dissolution, a fuel fraction that had previously been leached for over one year was reused. The U-238 and Tc-99 concentration was found to vary in the samples taken over 1100 days of leaching, depending on the degree of centrifugation. The erratic behavior of this autoclave experiment is tentatively attributed to a high surface area, with sub-micrometer sized fuel particles adhering to larger fuel fragments (evidenced by electron microscopy), caused by the fuel milling at the start of the experiment. This likely promoted an increased amount of pre-oxidation of the fuel as well as the potential for reductive precipitation and subsequent release of colloids from the autoclave. As a comparison, initial results from an ongoing autoclave experiment with coarser fuel fragments are also given.