Research has been carried out to optimize the consolidation stage for the immobilization of pyrochemical wastes with a sodium aluminophosphate glass. The alternative techniques of hot pressing and hot isostatic pressing of the calcined wastes with the glass have been investigated. This has been performed on simulant waste material and the products investigated by scanning electron microscopy and X‑ray diffraction. The consolidation techniques were compared to each other and to the original process for suitability as a waste-form.

Since its creation in 1950, the Argentine Atomic Energy Commission (CNEA), has worked on the development of applications for the peaceful use of nuclear energy. They include, among others, research and development activities in basic and nuclear technology areas, the operation of important facilities for the production of radioisotopes and the performance of tasks in connection with the nuclear fuel cycle, mining and uranium processing activities, manufacturing of fuel elements, production of heavy water and the operation of two nuclear power plants. Also demonstrating reprocessing programs have been developed.

As a result of the above mentioned activities performed in the nuclear field by private and public entities, various types of radioactive waste have been and are produced during operation and maintenance of facilities and plants, and will be produced in their future decommissioning. CNEA is also responsible for the environmental remediation of the closed uranium mining and milling facilities and for the management of disused sealed sources from medical and industrial applications.

Activities concerning radioactive waste management are carried on according not only to Nuclear Regulatory Authority requirements and International Atomic Energy Agency recommendations, but also to the national legal framework.

The Law Nº 25.018, "Radioactive Waste Management Regime“(1998), created the National Program for Radioactive Waste Management (PNGRR), as part of CNEA organization.

Spanish Radioactive Waste Management policy is established by the Government and implemented by ENRESA. The General Plan (GRWP) covers the analysis of the actual and foreseen inventories of spent fuel and all categories of radioactive waste, their present situation, the management strategy and actions identified, as well as funding and financial provisions. Very Low (VLLW) and Low and Intermediate Level Waste (LILW) are disposed of at El Cabril facility, which has two separate disposal areas: one intended for VLLW, based on clay and polyethylene and started up in 2006; and one for LILW conditioned in retrievable concrete containers, commissioned in 1992. Spent fuel (SF) is being stored in pools and in two dry storage installations. The priority is the development of the SF and HLW centralized storage facility. In 2009, the Government launched a call for candidate municipalities, in a public, participative process. In September 2010 a report was sent to the Cabinet, proposing eight volunteer communities, four of them deemed preferred. In the 90’s a deep geological repository (DGR) site identification program was carried out. DGR basic designs and associated performance assessments were developed in three rock types. ENRESA has set a research program that includes research projects in order to strengthen the link between management and scientific basis and improve the performance assessment, directed to the behavior of barriers, fuel or waste forms in different storage and final disposal conditions, and characterization techniques. There is also a participation in connected fields or supporting decision-making, such as advanced separation and transmutation.

Durable crystalline actinide host phases of ceramic waste forms are considered as advanced materials which are prospective for safe use of Pu and minor actinides before their final disposal. Development of self-glowing actinide-doped materials with matrices that are chemically inert and resistant to radiation damage may significantly change the approaches to actinide immobilization. Single crystals of zircon doped with different amount of Tb and 238Pu were synthesized by the flux method. Different non-radioactive crystals of Tb-doped zircon were studied first by cathodoluminescence method in order to identify the optimal content of Tb3+ that provides the highest luminescence emission. Then self-glowing crystals of zircon were grown with the optimal Tb content and small admixture of 238Pu (less than 0.1 wt. %). It was proposed that the valence state of Tb incorporated into zircon crystals can be (3+) and (4+), but only trivalent Tb is responsible for intensive luminescence. It is demonstrated that a small addition of Zr-phosphate to the flux supports Tb incorporation into zircon lattice and stabilizes preferably Tb3+. At the same time the addition of Zr-phosphate caused the crystallization of zirconia as a minor phase. Zircon crystals with very intensive self-glowing were successfully synthesized. The 238Pu content was 0.02 wt.% and the Tb concentration varied between 0.2 and 0.3 wt.%. Zirconia crystals obtained from the same experiment are characterized by weak self-glowing, although the Tb content was only 0.02 wt.%, while the content of 238Pu was comparable to that of zircon, i.e. 0.03 wt. %.

Calcium and barium zirconium phosphates were prepared by hot isostatic pressing and their thermophysical properties investigated for potential use as actinide hosts for inert matrix fuels (IMF) in light water reactors. The materials are thermally stable up to at least 1600°C in air, however they degrade above around 1400°C in an inert atmosphere. The heat capacity and thermal conductivity were measured from room temperature up to 1200°C. The thermal conductivity coefficient for both CZP and BZP at 1000°C is 1.0 W m-1 K-1, a relatively low thermal conductivity that requires NZP-type materials to be dispersed in a composite cercer or cermet IMF.

The pyrolysis of ion exchange resin beads that are used for the purification of water in reactor primary- and secondary-cooling circuits can result in stable and leach resistant carbonaceous products. However, free flowing beads are less desirable waste forms for disposal in sub-surface or surface repositories than monolithic masses of low porosity. We have investigated the pyrolysis of polymeric resin – cation exchange resin composites to give mechanically robust and chemically durable monolithic carbonaceous waste forms that are suitable for repository disposition. Also investigated was the dependence of product properties on various processing parameters (temperature ramp and final temperature). As a first approach, epoxy resins were used for the preparation of monoliths since such resins cure at room temperature and result in a relatively high carbon yield. Carbonaceous monolithic products were prepared at 400, 500, 600, 700 and 800 °C using a temperature ramp of 2°C/min. The products were maintained at the chosen temperatures for a period of one hour. Mass losses, volume reduction, hardness and compressive strength were measured and mathematical functions are proposed to describe the measured values of these properties. The carbon monoliths were observed to be mechanically robust.

Current energy demands and future energy needs are a growing industry which at present attracts a large amount of research and investment of which nuclear energy is an integral part. Eight new nuclear stations are proposed to be developed in the UK over the next ten years to meet this demand. In order for nuclear energy to sustain growth and development, nuclear decommissioning of first and second generation power stations needs to be addressed in the U.K. and worldwide. Presently the UK has 36 graphite moderated reactors as a result of the UK military and civil programs, which over the next twenty years will close. This will result in ∼99’000 tonnes of irradiated graphite waste for which no current national decommissioning strategy exists. The main issues associated with this waste are the large volume and activation products associated. By far the greatest inventory is from 3H and 14C. An EU Euroatom FP7 Program; CARBOWASTE was established in 2008 with the aim of developing treatment and disposal options for graphite.

This research is based within CARBOWASTE, the main objectives are to understand the mechanisms involved in the production, location and removal of radioisotopes from nuclear graphite. Computed X-ray Tomography (CT) will be used in order to quantify the initial porosity in conjunction with thermal treatment (ex situ) in order to eventually identify the location of 14C within the matrix of irradiated graphite, through the preferential chemically controlled oxidation of graphite. Unirradiated Pile Grade A graphite samples have been laser and manually marked in order align the samples prior to and post thermal treatment to determine the degree of porosity changes and weight loss under a range of thermal oxidation parameters.

The structural integrity of nuclear fuel cladding is affected by the precipitation of hydrides during operation, which may embrittle the cladding. The aim of this work is to obtain the mechanical and fracture properties of the cladding as a function of the hydrogen content and testing temperature. To this end, the embrittlement caused by circumferential hydrides was simulated on unirradiated fuel cladding samples in the laboratory. The structural integrity of the cladding was assessed at different temperatures (20, 135 and 300ºC), by using the ring compression test. The mechanical properties and the fracture energy were calculated from the experimental load vs. displacement curves, by means of a finite element model which incorporates the cohesive crack model.

To assess the long-term behavior of spent fuel in alkaline conditions representative for the Belgian Supercontainer design, static and dynamic dissolution tests were performed with depleted and Pu-doped UO2 , simulating medium burn-up UOX fuels of different fuel ages. The experiments were performed under argon atmosphere at 25 – 30°C in cement waters in the pH range 11.7 – 13.5 and at different SA/V ratios. This paper presents the observed UO2 matrix dissolution rates based on the (238U or 233U) release, and proposes a selection of reference dissolution rates for performance assessment. We demonstrate that the dissolution rates at high pH are equivalent to the dissolution rates reported in the literature for neutral pH conditions. The α-activity threshold below which radiolytical fuel oxidation becomes negligible, seems to be close to the threshold reported for anoxic media at neutral pH.

In the framework of long term prediction of corrosion in French geological repository systems, the modelling of the time evolution of the corrosion rate of iron over centuries is of high matter. The DPCM (Diffusion Poisson Coupled Model), implemented with the fully implicit CALIPSO numerical code can give access to the evolution of the oxide thickness grown on iron and the resulting corrosion rate. DPCM parameters for outer interfacial reactions have been set to fit to XANES experimental data provided by literature, and then the parameters for inner reaction rates have been evaluated from fitting to ellipsometry results taken from literature. The result is a good average fitting of the model to the experimental data.

Selection of appropriate sites for disposal of radioactive waste, especially high level waste and spent nuclear fuel, is a controversial task, not only from technological but also from societal point of view. A key part of the nuclear facility development is public consultation before the siting, construction and operation of the new repository. All decisions on these issues should be made in clear and transparent manner. The involvement of the local community from the very beginning of planning process may avoid faults and misunderstandings resulting in social objections and organized protests in future. To enhance the public participation in decision-making process several approaches of communication with the society were elaborated in the countries with well-developed nuclear power industry. Special models for communicating with stakeholders to build the acceptance and confidence concerning the radioactive waste management may be also helpful for Poland – the country entering the nuclear energy pathway. An effort to adapt the RISCOM Process developed in Sweden and also implemented in Czech Republic, to Polish conditions will be made in the scope of the EC-FP 7 IPPA project.

The stress corrosion cracking (SCC) of Alloy 22 in concentrated ground waters was associated to the presence of bicarbonate and chloride ions in the water. SCC occurred when an anodic peak appeared in the polarization curve of Alloy 22. The aim of this work was to investigate further which element in Alloy 22 was responsible for the anodic peak and therefore SCC. Four alloys (22, 800H, 600 and 201) were mainly used for this study. It has been found that Alloy 22, 600 and 800H show anodic peaks; which are affected by chloride and temperature. It is likely that the anodic peak is associated to the presence of either chromium, molybdenum or both.

Samples of AA 6061 alloy and Al 5N were anodized at constant potential conditions and sealed in distilled boiling water. Immersion tests in high purity water were also carried out on AA 6061 samples at open circuit potential. EIS measurements were then performed in order to investigate the properties of the oxides obtained.

For sealed anodic oxides on AA 6061 and Al 5N, EIS experimental data was useful to differentiate between the capacitances of the two layers present in the oxide, named the barrier and porous layers. EIS data obtained for oxide layers grown at open circuit potential in AA 6061 allowed estimating the capacitance of the barrier layer, which value was greater than the barrier layer capacitance present in anodic oxides.

Pyrochlores based on the general composition CaLnZrNbO7 (where Ln = La, Nd, Sm, Gd and Ho) have been prepared, and irradiated through the crystalline-amorphous transition, with 1 MeV Kr ions at the IVEM-TANDEM user facility. The obtained critical temperatures show a decrease from La to Gd (∼680 K to ∼230 K), with Ho being resistant to amorphisation at 50K. The results suggest that the amorphisation cross section for these materials is directly related to the Ln component.