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During irradiation UO2 nuclear fuel experiences the development of a non-uniform distribution of porosity which contributes to establish varying mechanical properties along the radius of the pellet. Radial variations of both porosity and elastic properties in high burnup UO2 pellet can be investigated via high frequency acoustic microscopy. For this purpose ultrasound waves are generated by a piezoelectric transducer and focused on the sample, after having travelled through a coupling liquid. The elastic properties of the material are related to the velocity of the generated Rayleigh surface wave ( \mathrm V \mathrm {R} ). A UO2 pellet with a burnup of 67 GWd/tU was characterized using the acoustic microscope installed in the hot cells of the JRC-ITU at a 90 MHz frequency, with methanol as coupling liquid. \mathrm V \mathrm {R} was measured at different radial positions. A good agreement was found, when comparing the porosity values obtained via acoustic microscopy with those determined using SEM image analysis, especially in the areas close to the centre. In addition, Young’s modulus was calculated and its radial profile was correlated to the corresponding burnup profile and to the hardness radial profile data obtained by Vickers micro-indentation [ABSTRACT FROM PUBLISHER]

Recent electron microscopy investigations are presented with an emphasis on properties related to the safe extended operation of nuclear fuel. In particular, this article considers the formation of the high burnup structure (HBS), the prevailing microstructural aspects associated to spent fuel aging, and the release of fission products under accidental condition. Examples of microstructures associated with transient-tested samples but also with very high burnup fuel or heterogeneous mixed oxides (MOX) are presented together with results on damage formation in UO samples doped with Pu to study the specific effect of alpha damage on the microstructure during the cooling/storage time of irradiated fuel. Examples of single-effect studies (e.g., on the behavior of tellurium, a typical volatile fission product) using ion implantations are also presented. [ABSTRACT FROM AUTHOR]

Abstract: A set of devices was developed in order to infuse UO2 disks with helium, at high temperature and pressure, to measure the helium infused quantity and from these data to calculate the helium solubility in the UO2 matrix. Samples of UO2 single crystal and UO2 polycrystal were infused at a temperature of 1473 and 1743K in a helium atmosphere ranging between 50 and 100MPa. These samples were then annealed and the helium released was measured with a mass spectrometer. From the obtained spectra it was possible to give an interpretation of the helium release mechanism and to calculate its solubility in the UO2 lattice in these specific thermodynamic conditions. Additionally to the helium solubility measurement from infused samples, a 37 years old sample of 238PuO2, retrieved from an old 242Cm radioisotope thermoelectric generator (RTG), containing radiogenic helium, was also measured to widen perspectives of this kind of measurements to damaged sample more representative of spent fuel. [Copyright &y& Elsevier]

Abstract: Mobilisation of uranium in geologic environments from UO2 solid phases usually takes place by oxidative dissolution involving a change of U oxidation state from +IV to +VI; however, anoxic or reducing geochemical conditions are expected in many of the planned European disposal sites. This work investigates potential alteration mechanisms of UO2 in contact with groundwater ions (Ca2+, , and silicate) under anoxic conditions, at ambient (25°C) and hydrothermal (180°C) temperature conditions. SEM–EDX analysis detected (in the case of treatment at 180°C in high silicate content solutions) a compound with U:Si ratio of 1:1 on the UO2 surfaces after leaching. Minor quantities of phases containing U, Ti, Fe, Si, and Ca were formed, these could not be characterized completely. A further experiment, performed in the presence of dissolving CaO/TiO2/SiO2/Fe(0)–Fe2O3, formed a compound with U:Si:Ca of 1:2:8, a ratio not matching any known uranyl compound. The two phases, possibly identical with coffinite, USiO4, and U-bearing ekanite, UCa2Si8O20, were found to form at different [Ca]/[Si] conditions. The implications upon the final geologic storage of actual, heterogeneous spent fuel are discussed. [Copyright &y& Elsevier]

UO 2 samples doped with the strong α-emitter 238Pu were periodically characterised to study the potential effects of α-self-irradiation on spent nuclear fuel during storage. The degradation of the thermal properties and microstructure of the samples were monitored using the Laser Flash technique (LAF), X-Ray Diffraction (XRD), and Scanning and Transmission Electron Microscopy (SEM, TEM). The thermal conductivity decreased with a non-linear behaviour as a function of the α-dose. Three clearly different stages in the thermal conductivity degradation could be detected, and saturation was reached at relatively low doses (0.035 dpa). In combination with analyses of the microstrain measured by XRD, and TEM images, these three phases could be attributed to three stages of radiation-induced defect generation and clustering. SEM inspection showed the thermal conductivity degradation cannot be due to loss of integrity or grain boundary opening. The radiation damage produced resulted in the formation of dislocation loops as observed by TEM. [ABSTRACT FROM AUTHOR]

Scanning acoustic microscopy is a non-destructive technique that allows determining the local material elastic properties by measuring the velocity of acoustic waves propagating in matter. High frequency acoustic waves are generated by a piezoelectric transducer, focused and then detected by the same transducer after having interacted with the sample. This technique has been employed in the past to assess different types of irradiated nuclear fuel and unirradiated chemical analogues of UO2 and it has enabled to relate the Rayleigh wave velocity of propagation with the Young's modulus and the density of the material. In the present study, thanks to new measurements on irradiated fuel and to analysis of data from the open literature, the variation of the density with burnup is determined up to ∼ 100 GWd·t −1 M. The porosity is then determined taking account of the irradiated fuel matrix swelling. Finally, an expression is proposed describing Young's modulus as a function of burnup, that can be used in fuel performance calculation. [ABSTRACT FROM AUTHOR]

During reactor operation the fission gases Kr and Xe are formed within the UO 2 matrix of nuclear fuel. Their quantification is important to evaluate their impact on critical parameters regarding the fuel behaviour during irradiation and (long-term) interim storage, such as internal pressure of the fuel rod and fuel swelling. Moreover the content of Kr and Xe in the plenum of a fuel rod and their content in the UO 2 fuel itself are widely used as indicators for the release properties of 129 I, 137 Cs, and other safety relevant radionuclides with respect to final disposal of spent nuclear fuel. The present study deals with the fission gas release from spent nuclear fuel exposed to simulated groundwater in comparison with the fission gas previously released to the fuel rod plenum during irradiation in reactor. In a unique approach we determined both the Kr and Xe inventories in the plenum by means of a puncturing test and in leaching experiments with a cladded fuel pellet and fuel fragments in bicarbonate water under 3.2 bar H 2 overpressure. The fractional inventory of the fission gases released during irradiation into the plenum was (8.3 ± 0.9) %. The fraction of inventory of fission gases released during the leaching experiments was (17 ± 2) % after 333 days of leaching of the cladded pellet and (25 ± 2) % after 447 days of leaching of the fuel fragments, respectively. The relatively high release of fission gases in the experiment with fuel fragments was caused by the increased accessibility of water to the Kr and Xe occluded in the fuel. [ABSTRACT FROM AUTHOR]

Abstract: Green rust (GR), a member of the Fe(II),Fe(III) layered double hydroxide mineral family, forms in groundwater and during steel corrosion. It has high surface area and is very reactive, especially for redox-sensitive elements such as some actinides. During neutron irradiation of nuclear fuel in a reactor, 237Np develops. Although the abundance of Np in spent nuclear fuel is only about 0.05% by mass, it has a very long half life, 2.14×106 years, so there is concern about its mobility in the distant future, when radioactive storage sites might be expected to degrade. Under oxidizing conditions, as are expected from radiolysis in a repository, the very mobile neptunyl-ion, Np, would be the dominant aqueous neptunium species. In this work, we investigated the interaction of Np with green rust sodium sulphate (). The aim of the study was to define the processes involved, to determine the final redox speciation of Np, hence its potential mobility, and to characterise changes in the green rust. The sorbed and reduced Np within minutes. Reduced Np(IV) was primarily found as precipitated nanoparticles at the edges of the crystal platelets. The position of the particles at the crystal edges suggests initial sorption of Np(V) and subsequent reduction at GR edge sites. In further experiments, Np-containing GR was allowed to oxidise completely in aqueous suspension. Surprisingly, about 50% of the Np remained associated with the oxidised product and about 75% of the associated Np was still present as Np(IV). [Copyright &y& Elsevier]

Abstract: The presented work aims to reproducibly prepare UO2–Pd thin film model systems for spent nuclear fuel in order to further investigate surface reactions of these films under relevant redox conditions. The sputter co-deposition of U and Pd in the presence of O2 results in the homogeneous distribution of Pd in a crystalline UO2 matrix. Hereby, Pd is found to be oxidized and to form PdO x . Heating the films after deposition causes the diffusion of film components and induces a change in surface morphology. Independent of the heating temperature initial UO2+ x transforms into UO2. This is different for the noble metal. At high temperatures (550–840°C) Pd diffuses into the Si-wafer substrate and forms mixed Pd–Si–U alloys. At moderate temperatures (150–200°C) Pd solely diffuses within the film matrix and forms micrometer sized metallic particles. These particles are further characterized as being an agglomeration of small nanometer sized spheres. [Copyright &y& Elsevier]

Abstract: The presented work aims to reproducibly prepare UO2–Pd thin film model systems for spent nuclear fuel in order to further investigate surface reactions of these films under relevant redox conditions. The sputter co-deposition of U and Pd (fission product) in the presence of O2 results in the homogenous distribution of Pd in a crystalline UO2 matrix. Heating the films causes the diffusion of film components. Hereby, the formation of ε-particles has to be clarified. First electrochemical studies show the influence of the nobel metal Pd on the redox behaviour of UO2. With increasing Pd concentration the matrix dissolution is decreased. However, we could demonstrate that blocked oxidation processes are of temporary nature. The passivation of the Pd reactive sites with increasing number of cycles finally induces the approximation of the mixed system to the redox behaviour of the pure UO2 system. [Copyright &y& Elsevier]

Zirconates with the pyrochlore structure (A2Zr2O7) are investigated at ITU for use as an actinide host in inert matrix fuels for transmutation. Zirconate pyrochlores with A=Nd as an inactive stand in for the trivalent actinides Am and Cm were fabricated and characterised, and their thermal transport properties measured. The low thermal conductivity indicates that zirconate pyrochlore can only be used for transmutation if it is dispersed in a cercer or cermet composite fuel. The temperature profiles of a MgO–An2Zr2O7 composite were calculated using the measured Nd-zirconate thermal conductivity for different concentrations of the included phase. The radiation stability of Nd2Zr2O7 against fission products (FP) was investigated using iodine irradiation (120 MeV). Significant alterations of the implanted regions were observed even at relatively low fluence. [Copyright &y& Elsevier]