Enhanced thermal conductivity uranium dioxide composites containing silicon carbide (UO2−SiC) and diamond (UO2-diamond) have been irradiated to low burnup. The conditions of this irradiation test and ...subsequent postirradiation examinations are discussed. These irradiations evaluate fuel microstructure and potential fuel cladding interaction of UO2 composites, which have been proposed as accident tolerant fuel candidates.
Both non-destructive and destructive techniques have been used to evaluate fuel integrity, fission gas release, fission product distribution, burnup, fuel swelling and cladding strain. Examination of the UO2-SiC pellets revealed enhanced cracking when compared to UO2 pellets irradiated under similar conditions. Instability of the SiC whiskers in the uranium dioxide matrix was observed in the pellet central region, where the local temperatures exceeded 1300°C. The microstructure of the UO2-diamond was severely disrupted during irradiation, resulting in local migration of cesium along the fuel stack and increased fission gas release when compared with the expected release from the Vitanza curve at corresponding values of burnup and irradiation temperature.
The postirradiation examination results cast doubt on the suitability of these additives to improve UO2 fuel performance in a way that would lead to enhanced accident tolerance.
We propose a model describing the HBS formation and the progressive intra-granular xenon depletion in UO2. The HBS formation is modeled employing the Kolmogorov-Johnson-Mehl-Avrami (KJMA) formalism ...for phase transformations, which has been fitted to experimental data on the restructured volumetric fraction as a function of the local effective burnup. To this end, we employed available experimental data and novel data extracted in this work. The HBS formation model is coupled to a description of the intra-granular fission gas behavior, allowing to estimate the evolution of the retained xenon in order to consistently compute fission gas retention and its effect on the fuel matrix swelling. The satisfactory agreement of the model predictions to experimental data and state-of-the-art models’ results, in terms of both xenon depletion and fuel matrix swelling as a function of the local burnup, paves the way to the inclusion of the model in fuel performance codes.
We present post-irradiation examination results on two type of annular mixed oxide fuel pins irradiated in the Fast Flux Test Facility (FFTF) sodium cooled reactor to an average burnup between 4% and ...5% fission of initial heavy atom (FIMA). The pins differed only from the initial Pu content, which was 22 wt% and 26 wt%, respectively. The overall performance of the pins was excellent, in line with previous historical results. The pins with higher Pu content experienced higher irradiation temperatures which influenced the fission gas release, fuel swelling, and Cs distribution compared to the other pins. All the post-irradiation examinations results are discussed against the irradiation parameters. In particular, the pins with higher initial Pu content, i.e., 26 wt%, experienced higher power that resulted in enhanced fission gas release compared to the other two pins with 22 wt% initial Pu content. For the pins with higher fission gas release, onset of Cs redistribution was observed. The two pins that had lower initial Pu content and burnup showed a Cs axial distribution similar to the as-produced one.
Continued research into the mechanical properties of nuclear fuel is necessary to improve modeling of the pellet-cladding mechanical interactions that occur in nuclear reactors during operation. ...Small scale mechanical testing with focused ion beam equipment can provide a means to study the mechanical properties of irradiated nuclear fuels by measuring their mechanical properties on minute volumes of material. In this work, room temperature nanoindentation measured the hardness and Young's modulus of fast reactor mixed oxide (MOX) fuel irradiated to 183 GWd/tHM. The hardness of the MOX fuel had a measured value of 11.2 ± 0.4 GPa. The Young's modulus of the MOX fuel had a value of 158 ± 6 GPa. The measured hardness and Young's modulus are discussed in the context of the available literature data. In addition, pathways toward elevated temperature testing and other mechanical properties are proposed.
This paper introduces a method to reconstruct the three-dimensional (3D) microstructure of two-phase materials, e.g., porous materials such as highly irradiated nuclear fuel, from two-dimensional ...(2D) sections via a multi-objective optimization genetic algorithm. The optimization is based on the comparison between the reference and reconstructed 2D sections on specific target properties, i.e., 2D pore number, and mean value and standard deviation of the pore-size distribution. This represents a multi-objective fitness function subject to weaker hypotheses compared to state-of-the-art methods based on n-points correlations, allowing for a broader range of application. The effectiveness of the proposed method is demonstrated on synthetic data and compared with state-of-the-art methods adopting a fitness based on 2D correlations. The method here developed can be used as a cost-effective tool to reconstruct the pore structure in highly irradiated materials using 2D experimental data.
In this work, we present electron microscopy data focused on the fuel-cladding interaction layer in annular fast reactor MOX with HT-9 cladding at medium burnup. In agreement with previous literature ...data, the volatile fission products Cs, Te and I have migrated radially into the extreme fuel periphery and partially interacted with the cladding. The accumulation of Cs has occurred in the outermost rim of the fuel pellet, where grain recrystallization has also been observed. Significant amounts of Pd have been found in the interaction zone, particularly in the sample taken from the upper half of the fissile column where the cladding temperatures are higher. At this axial location, Cr has been enriched at the cladding inner surface and diffused into the fuel. Chromium remains mainly in metallic form, but locally formed oxides. The fission products Cs, Te and I are found, with variable composition, in form of nanocrystalline regions dispersed in the metallic Cr-rich layer. The morphology and chemical characteristics of the layer suggest a non-oxidative corrosion mechanism as principal cladding degradation phenomenon occurring in this sample, with local onset of Cr oxidation within the nanocrystalline precipitates.
This study presents the first thermal conductivity measurements performed on neutron irradiated U3Si2 dispersion fuel cladded in an aluminum matrix (in a plate geometry). These were performed via a ...novel set-up which consists of a thermo-reflectance apparatus, known as the thermal conductivity microscope (TCM), deployed in a shielded glovebox. Thermal conductivity was measured across different fuel particles at various positions along the plate length and correlated to the local burnup. Back scattered electron images were collected at each position and analyzed to obtain estimates of the local porosity. It was shown that the thermal conductivity of the irradiated fuel particles is approximately 25% to 35% lower compared to the unirradiated material. A solid-state physics model was used to interpret the measured data. According to the model the degradation in thermal conductivity is consistent with the presence of insulating, fission gas filled pores or bubbles as well as loss of long-range order due to irradiation induced amorphization.
The Advanced Fuels Campaign performed a series of irradiation tests of minor actinide-bearing mixed oxide fuel (MA-MOX), the so-called AFC-2C&D experiments, to investigate the transmutation of ...long-lived transuranic actinide isotopes contained in spent nuclear fuel via fast reactor technology at burnups exceeding 10 % fission of initial metallic atoms. This manuscript reports the test results derived from one of the five MA-MOX rodlets taken to higher burnup in the AFC-2D irradiation. This includes both non-destructive investigations, such as gamma and neutron spectrometry, and destructive investigations, such as fission gas release, ceramography, and chemical burnup analysis. In addition, the microstructure of the fuel was investigated using advanced electron microscopy techniques including electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). It was observed with EBSD that the pellet had subdivision of the grains and the TEM observed migration of cladding material into the 5 metal precipitates in the fuel which could have been from the higher than desired oxygen/metal ratio. The TEM also showed an enrichment of Cr in fuel clad chemical interaction (FCCI) layer.
The composition and crystal structure of the “Joint Oxyde Gaine” (JOG) has been investigated by means of electron microscopy. Microstructural characterization reveals a highly heterogeneous porous ...structure with inclusions containing both fission products and cladding components. Major fission products detected, other than Cs and Mo, are Te, I, Zr and Ba. The layer is composed by sub-micrometric crystallites. The diffraction data refinement, together with chemical mapping, confirms the presence of Cs2MoO4, which is the major component of the JOG. However, combinatorial analyses reveal that other non-stoichiometric phases are possible, highlighting the complex nature of the crystalline structure of the JOG.
Fe is found in metallic Pd-rich precipitates with structure compatible with the tetragonal structure of FePd alloy. Cr is found in different locations of the JOG, in oxide form, but no structural data could be obtained due to local beam sensitization of the sample in those areas.
This work applies reconstruction methods based on a genetic algorithm to derive 3D material properties, namely porosity and percolation fraction, in irradiated U-Pu-Zr fuel with minor actinides. We ...provide two-dimensional experimental data regarding the radial distribution of fission gas bubbles in the fuel and apply the algorithm successfully developed in a companion paper to reconstruct the fuel pore structure in 3D which is unknown a priori. The algorithm returned a set of best structures that constituted the best candidate solutions representing the pore phase. From these, it was possible to extract statistics on the 3D percolation fraction of the reference medium and infer a mean value, the related uncertainty, and an upper and lower bound of the percolation fraction. The algorithm proved able to infer this 3D property from 2D information of the metallic fuel with confidence intervals, thus establishing a path to infer 3D properties directly from 2D experimental images. The knowledge of such a relationship can be used to extrapolate the percolation threshold with confidence interval, which is a crucial property in defining microstructure-based fission gas release models of metallic fuels.
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