SUPERFACT: A Model Fuel for Studying the Evolution of the Microstructure of Spent Nuclear Fuel during Storage/Disposal

The transmutation of minor actinides (in particular, Np and Am), which are among the main contributors to spent fuel α-radiotoxicity, was studied in the SUPERFACT irradiation. Several types of transmutation UO$_{2}$-based fuels were produced, differing by their minor actinide content ($^{241}$Am, $^{237}$Np, Pu), and irradiated in the Phénix fast reactor. Due to the high content in rather short-lived alpha-decaying actinides, both the archive, but also the irradiated fuels, cumulated an alpha dose during a laboratory time scale, which is comparable to that of standard LWR fuels during centuries/millenaries of storage. Transmission Electron Microscopy was performed to assess the evolution of the microstructure of the SUPERFACT archive and irradiated fuel. This was compared to conventional irradiated spent fuel (i.e., after years of storage) and to other $^{238}$Pu-doped UO$_{2}$ for which the equivalent storage time would span over centuries. It could be shown that the microstructure of these fluorites does not degrade significantly from low to very high alpha-damage doses, and that helium bubbles precipitate

Comparative study of fission product release of homogeneous and heterogeneous high-burn up MOX fuel by Knudsen Effusion Mass Spectrometry supported by EPMA, SEM/TEM and thermal diffusivity investigations

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The high burnup structure in nuclear fuel

During its operating life in the core of a nuclear reactor, nuclear fuel is subjected to significant restructuring processes determined by neutron irradiation directly through nuclear reactions and indirectly through the thermo-mechanical conditions established as a consequence of such reactions. In today's light water reactors, starting after ~4 years of operation the cylindrical UO2 fuel pellet undergoes a transformation affecting its outermost radial region. The discovery of a newly forming structure required answering important questions concerning the safety of extended fuel operation, and still today poses a fascinating scientific challenge to fully understand the microstructural mechanisms responsible for its formation.JRC.DG.E.2-Hot cell

Leaching of SIMFUEL in Granitic Water. Comparison to Results in Demineralized Water

Statis leaching of SIMFUEL in granite water at 200*C for 1, 10, 100, and 1000 hours under argon atmosphere and in presence of granite monoliths was studied. Three simulated burnup compositions (3, 6 and 8 at.%) were used. The results did not indicate any clear burnup effect. These results were compared to results obtaineds from leaching tests on SIMFUEL and UO2 in demineralized water in air. Under these conditions, leach rates were significantly higher. With increasing leaching time, the surfaces became increasingly covered by crystals of redeposited uranium in an higher oxide form (e.g. schoepite).JRC.E-Institute for Transuranium Elements (Karlsruhe

Construction of a gas-mixing and analysis facility for fission gas release and analysis studies.

The formation of fission gases in the fuel during operation and their rate of release to the grain boundaries and hence to the plenum is a major feature that can limit the linear power rating of the fuel and the duration of a reactor’s continuous operation. Much attention has been given to investigating the mechanisms of fission gas accumulation in the fuel matrix and its gradual collection into bubbles, both within the grains and at grain boundaries and how to maximize its retention in the fuel matrix. Considerable work has been done at JRC-ITU and elsewhere looking at the mechanisms and irradiated fuel history effects on fission gas retention/release from the fuel. The current concern for commercial reactors is particularly about fission gas behaviour for MOX or high burn-up UO2 under operating power transients. Gradual accumulation in the rod free volume would increase the internal rod pressure and may ultimately induce cladding rupture. In extreme cases, such as power transients or accident conditions, sudden fission gas release can result in multiple rods failure and release of fission gas and other volatile fission products to the primary water circuit, forcing reactor shut down. This paper will look at the construction and preliminary testing at JRC-ITU of a gas-mixing facility and its analytical devices (mass spectrometer, cold traps and gamma spectrometers) along with its connection to a high temperature furnace in view of new campaigns to assess fission gas behaviour during off-normal conditions from commercial or innovative nuclear fuels.JRC.G.III.8-Waste Managemen

POTENTIAL STRESS ON CLADDING IMPOSED BY THE MATRIX SWELLING FROM ALPHA DECAY IN HIGH BURNUP SPENT NUCLEAR FUEL

Studies using literature data were conducted on potential swelling of the spent nuclear fuel (SNF) matrix due to alpha decays. Swelling of the matrix may impose stress on cladding during long-term storage of SNF. The literature data were obtained for UO2 having undergone accelerated irradiation conditions. The representation of the simulated accelerated tests for the SNF is discussed. The stress on cladding, and its potential implications in terms of long-term degradation, would depend on SNF matrix swelling, gap size, crack pattern and crack interlocking, and friction of the SNF matrix and cladding. Assessments and reviews are made for the observed SNF matrix swelling, stress generated on cladding, and gap size. Applying this information, the stress on cladding due to alpha decays is semi-quantitatively assessed using stress analysis induced by thermal expansion.JRC.E.2-Hot cell

Comparison of fluorescence-enhancing reagents and optimization of laser fluorimetric technique for the determination of dissolved uranium

Results from tests aimed at optimizing an instrumental procedure for the direct and fast determination of uranium in solution by laser fluorescence are presented. A comparison of sample fluorescence measured using different fluorescence enhancing reagents was performed: sodium pyrophosphate, orthophosphoric acid, sulphuric acid and a commercially available fluorescence enhancer were tested for the determination of uranium. From the experimental results, 0.01 M Na4P2O7-10H2O showed the best performance. Effects of reagent pH, different matrices, different concentrations of dissolved Th, and sample volume were investigated. Applications of the improved procedure for the determination of uranium in samples arising from UO2-based high level nuclear waste dissolution studies are described. © 2010 Akadémiai Kiadó, Budapest, Hungary.Peer Reviewe