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The Knudsen Effusion Mass Spectrometer (KEMS) and the mechanistic MFPR (Module for Fission Product Release) code are tools which seem particularly interesting to support studies of the Instant Release Fraction (IRF) of Cs from spent nuclear fuel in a final repository. With KEMS, the thermal release of 137Cs and 136Xe were analysed by annealing up to total vaporization (2500K) of high burn-up (60 GWd/tU) Spent Nuclear Fuel (SNF) samples. Powder samples from the centre of the fuel, without high burn-up structure, were used. To determine the IRF, samples were analysed before and after being submitted to corrosion experiments in bicarbonated aqueous media.MFPR was applied to determine the localization of Cs and fission gases in the SNF at the end of irradiation; the results are compared and supported by dedicated thermodynamics calculations performed for equilibrium conditions at various temperatures and fuel oxygen potentials by the non-ideal thermodynamic MEPHISTA (Multiphase Equilibria in Fuels via Standard Thermodynamic Analysis) database. A possible mechanism for Cs release during thermal annealing is proposed, taking into account inter-granular release and Cs oxide vaporization, atomic diffusion, ternary oxide phase formation and bubble release.Differences in KEMS release profiles before and after submitting the samples to aqueous corrosion are attributed to the IRF and to changes in the vaporisation mechanism because of differences in the oxygen potential (pO2). The IRF of Cs estimated from the KEMS spectra, consisting on the part located at the grain boundaries and in inter-granular bubbles, is not significantly different from that corresponding to the experimental results found using classical static leaching experiments.New experimental campaigns are being designed to confirm our interpretation proposed after this first run.

The dissolution of spent nuclear fuel is defined in two different time steps, i) the Instant Release Fraction (IRF) occurring shortly after water contacts the solid spent fuel and responsible of the fast release of those radionuclides that have been accumulated in the zones of the spent fuel pellet with low confinement, such as gap and grain boundaries and ii) the long term release of radionuclides confined in the spent fuel matrix, much slower and dependent on the conditions of the water that contacts the spent fuel.Several models have been developed to date to explain the dissolution behavior of spent nuclear fuel under disposal conditions. The Matrix Alteration Model (MAM) is one of the most evolved radiolytic models describing the dissolution mechanism in which an Alteration/Dissolution source term model is based on the oxidative dissolution of spent fuel. Under deep repository conditions and at the expected of water contacting time (after 1000 years of spent fuel storage), α radiation will be the main contributor to water radiolysis. In the current study, simulations evaluating the effect of surface area on the alteration/dissolution of spent fuel matrix are performed considering different particle sizes of spent fuel and simulations integrating the actinides dissolution have been performed considering the precipitation of secondary phases.

The assessment of the main changes expected for spent nuclear fuel from its discharge to its deposition in a deep geological repository is of the outmost relevance to establish the initial conditions of the disposal. In this work, a literature review and a critical discussion of the main processes that will affect the structure and the inventory of the spent nuclear fuel during its interim dry storage is presented. Once the irradiation period is finished, the following changes are observed: i) the fuel pellet is fragmented due to the temperature gradient established during the irradiation stage. On average between 10-15 fragments are observed per pellet. ii) the initial gap existing between the pellet and the cladding decreases or disappears depending on the burnup. iii) a radial zonation is observed in the microstructure of the pellet. For burnup over 40MWd/KgU, the rim develops a porosity increase due to the high local burnup and the low temperature in the periphery. The rim also presents small bubbles of fission gases. This high burnup structure implies a degradation of the thermic conductivity in the pellet, that leads to a temperature increase in the center of the pellet with a subsequent migration of the fission gases and other impurities to the grain boundaries. The implications that all these changes may have on the spent fuel behaviour is presented and discussed.

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.

European Commission supported a wide research project entitled “Spent Fuel Stability under repository conditions” (SFS) within the 5th FWP, the aim of which was to develop a common understanding of the radionuclides release from spent nuclear fuel in geological disposal and build a RN release model in order to assess the fuel performance. This project achieved by the end of 2004 focuses both on the Instant Release Fraction (IRF) model and the Matrix Alteration Model (MAM).A new IRF model was developed based on the anticipated performances of the various fuel microstructures (gap, rim, grains boundaries) and the potential diffusion of RN before the canister breaching. However, this model lets the choice to the end-user about the degree of conservativeness to consider.In addition, fuel alteration has been demonstrated to be linked to the production of radiolytic oxidants by water radiolysis at the fuel interface, the oxidation of the fuel interface by radiolytic oxidants and the subsequent release of uranium under the influence of aqueous ligands. A large set of experimental data was therefore acquired in order (i) to upgrade the current radiolytic kinetic scheme, (ii) to experimentally correlate the fuel alteration rate and the fuel specific alpha activity by performing experiments on alpha doped samples, (iii) to experimentally test the potential inhibitor effect of hydrogen on fuel dissolution. Based on these results, a new MAM was developed, which was also calibrated using the experiments on inactive UO2 samples. This model was finally applied to representative granitic, salt and clayey environment to predict spent fuel long-term fuel performance.

In this work we have developed a model for the release of radionuclides from the spent fuel coupled with their transport through the near field. A compartmental approach has been used, as this methodology is well suited to model integrated systems. Several processes have been taken into account: oxidative dissolution of the spent fuel matrix, radioactive decay and chains, diffusive and advective transport, retardation by sorption and secondary phase precipitation. Results illustrate the complex evolution of the radionuclide concentrations in the gap and the near field. Hence, the main conclusion from this study is the requirement to model this coupled system using a compartmental integrated approach.

One of the factors considered within the studies of performance assessment on spent fuel under final repository conditions is the effect of the radiation on its leaching behavior. Radiation from spent fuel can modify some properties of both solid phase and leachant and therefore it would alter the chemical behavior of the near field. Particularizing in the effect of the radiation on the leachant, it will cause generation of radiolytic species that could change the redox potential of the environment and therefore may bring on variations in the leaching process. In this work, the chemical analogue utilized was SIMFUEL (natural UO 2 doped with non-radioactive elements simulating fission products) and the leachants selected were saline and granite bentonite waters both under initial anoxic conditions. To emulate γ radiation field of a spent fuel, leaching experiments with external 60 Co sources in a irradiation facility (Nayade) were performed. Initial dose rate used was 0.014 Gy/s. Preliminary results indicate that radiation produces an increase of the uranium dissolution rate, being the concentrations measured close to those obtained in oxic atmosphere without radiation field. In addition the solubility solid phases from experimental conditions were calculated, for both granite bentonite water and 5 m NaCl media. On the other hand, a tentative approach to model the role of γ radiolysis in these SIMFUEL tests has been carried out as well.