D. Papaioannou

Lattice Parameter Changes Associated with the Rim-Structure Formation in High Burn-up UO2-Fuels by Micro X-Ray Diffraction.

Abstract Radial variations of the lattice parameter and peak width of two high burn-up UO 2 -fuels (67 and 80 GWd/tM) were measured by a specially developed micro-X-ray diffraction technique, allowing spectra acquisition with 30 μm spatial resolution. The results showed a significant but constant peak broadening, and a lattice parameter that increased towards the pellet edge and decreased again within the rim-zone. This lattice contraction coincided with other property changes in the rim region, i.e., porosity increase, hardness decrease and Xe depletion. In terms of local burn-ups, the lattice contraction followed the rate of the matrix Xe depletion measured by EMPA, exceeding greatly the contraction rate due to dissolved fission products. The observed behaviour can be equally explained by a saturation of single interstitials with subsequent recombination with excess vacancies, as by the saturation and enlargement of dislocation loops. The concentration and sizes of defects involved and their possible relation to the rim structure formation are discussed.

Development of Fast Reactor Metal Fuels Containing Minor Actinides

Fast reactor metal fuels containing minor actinides (MAs) Np, Am, and Cm and rare earths (REs) Y, Nd, Ce, and Gd are being developed by the Central Research Institute of Electric Power Industry (CRIEPI) in collaboration with the Institute for Transuranium Elements (ITU) in the METAPHIX project. The basic properties of U-Pu-Zr alloys containing MA (and RE) were characterized by performing ex-reactor experiments. On the basis of the results, test fuel pins including U-Pu-Zr-MA(-RE) alloy ingots in parts of the fuel stack were fabricated and irradiated up to a maximum burnup of ∼10 at% in the Phénix fast reactor (France). Nondestructive postirradiation tests confirmed that no significant damage to the fuel pins occurred. At present, detailed destructive postirradiation examinations are being carried out at ITU.

Reducing uncertainties affecting the assessment of the long-term corrosion behavior of spent nuclear fuel.

Reducing the uncertainties associated with extrapolation to very long term of corrosion data obtainable from laboratory tests on a relatively young spent nuclear fuel is a formidable challenge. In a geologic repository, spent nuclear fuel may come in contact with water tens or hundreds of thousands of years after repository closure. The corrosion behavior will depend on the fuel properties and on the conditions characterizing the near field surrounding the spent fuel at the time of water contact. This paper summarizes the main conclusions drawn from multiyear experimental campaigns performed at JRC-ITU to study corrosion behavior and radionuclide release from spent light water reactor fuel. The radionuclide release from the central region of a fuel pellet is higher than that from the radial periphery, in spite of the higher burnup and the corresponding structural modifications occurring at the pellet rim during irradiation. Studies on the extent and time boundaries of the radiolytic enhancement of the spent fuel corrosion rate indicate that after tens or hundreds of thousands of years have elapsed, very small or no contribution to the enhanced corrosion rate has to be expected from α radiolysis. A beneficial effect inhibiting spent fuel corrosion due to the hydrogen overpressure generated in the near field by iron corrosion is confirmed. The results obtained so far point toward a benign picture describing spent fuel corrosion in a deep geologic repository. More work is ongoing to further reduce uncertainties and to obtain a full description of the expected corrosion behavior of spent fuel.

Irradiation of Minor Actinide–Bearing Uranium-Plutonium-Zirconium Alloys up to ˜2.5 at. %, ˜7 at. %, and ˜10 at. % Burnups

An irradiation experiment on minor actinide (MA)-bearing uranium-plutonium-zirconium (U-Pu-Zr) alloys, in which contamination by rare earth (RE) elements was considered, was performed up to ˜2.5 at. %, ˜7 at. %, and ˜10 at. % burnups in the Phenix fast reactor. All the irradiated metal fuel pins were subjected to nondestructive tests such as cladding profilometry and gamma spectroscopy. Then, cross-sectional metallography of the low-burnup and medium-burnup fuel alloys was performed, and the redistribution of the fuel matrix constituents”U, Pu, and Zr”in the low-burnup fuels was analyzed by energy dispersive X-ray spectroscopy. As a result, the irradiation growth of MA-rich and RE-rich precipitates was observed by comparing the low-burnup and medium-burnup fuels. From the postirradiation examinations carried out so far, it was confirmed that the irradiation swelling, the cross-sectional structures, and the migration of matrix constituent in metal fuels containing 5 wt% or less MAs and REs are almost the same as those in conventional U-Pu-Zr fuels.

Influence of low-temperature air oxidation on the dissolution behaviour of high burn-up LWR spent fuel

Abstract High burn-up (53.1 GWd/t(U)) light water reactor (LWR) spent fuel (UO 2 ), previously oxidised in flowing air to O / M ratios varying from 2.0 to 2.44 (U 4 O 9+ x ), was sequentially leached in deionised water at room temperature. A clear increase in the leaching rate of the matrix and fission products with a high redox sensitivity, like technetium and molybdenum, was observed as a function of the degree of oxidation. This effect is limited to the initial step of leaching and was not observed for plutonium and for subsequent leaching periods, no relevant influence of the O / M ratio on the leaching rate could be detected. For the uranium matrix, a very low leaching rate ( 3×10 −7 g cm −2 d −1 ) was measured independent of the O / M ratio. In comparison to the matrix, all the fission products have higher release rates, suggesting a non-congruent leaching of elements segregated to the grain boundaries.

A microbeam collimator for high resolution x-ray diffraction investigations with conventional diffractometers

A collimating system has been developed for condensing hard x rays, providing very thin, intense, and low divergent beams. The primary radiation is compressed down to the micrometer scale by multiple total reflections in the channel between two oblong and flexible metallic mirrors with exit aperture 4 mm×15 μm. The flexible mirrors permit variation of the channel profile and opening for the incoming radiation, adjusted for maximum transmitted x-ray intensities. The condenser, due to the high brilliance gain of the obtained beam compared to the uncompressed radiation going through a slit of the same size, can be operated even with conventional x-ray tubes, e.g., common x-ray diffractometers without the need for expensive high intensity synchrotron radiation sources, as demanded usually by the glass monocapillaries. A prototype, being mounted on a commercial theta–theta diffractometer, has been thoroughly tested for intensity gain, divergence, and spatial resolution and utilized for accurate structure deter...

Properties of the high burnup structure in nuclear light water reactor fuel

Abstract The formation of the high burnup structure (HBS) is possibly the most significant example of the restructuring processes affecting commercial nuclear fuel in-pile. The HBS forms at the relatively cold outer rim of the fuel pellet, where the local burnup is 2–3 times higher than the average pellet burnup, under the combined effects of irradiation and thermo-mechanical conditions determined by the power regime and the fuel rod configuration. The main features of the transformation are the subdivision of the original fuel grains into new sub-micron grains, the relocation of the fission gas into newly formed intergranular pores, and the absence of large concentrations of extended defects in the fuel matrix inside the subdivided grains. The characterization of the newly formed structure and its impact on thermo-physical or mechanical properties is a key requirement to ensure that high burnup fuel operates within the safety margins. This paper presents a synthesis of the main findings from extensive studies performed at JRC-Karlsruhe during the last 25 years to determine properties and behaviour of the HBS. In particular, microstructural features, thermal transport, fission gas behaviour, and thermo-mechanical properties of the HBS will be discussed. The main conclusion of the experimental studies is that the HBS does not compromise the safety of nuclear fuel during normal operations.

Microbeam analysis of irradiated nuclear fuel

Microbeam analysis is widely used in the nuclear power industry. It is used to characterise the as-fabricated fuel, for routine post-irradiated examination and for research into the mechanisms of phenomena that limit the energy production of the fuel. The techniques most commonly used are wavelength-dispersive electron probe microanalysis, scanning electron microscopy and secondary ion mass spectrometry. Other microbeam analysis techniques that have been successfully applied to irradiated nuclear fuel are transmission and replica electron microscopy, X-ray fluorescence and micro X.-ray diffraction. Specific examples illustrating the past and present use of microbeam analysis in nuclear research establishments are presented with emphasis on the unique results they provide. As an aid to understanding, some basic facts about nuclear fuel rods and their irradiation are first given. This is followed by a description of features that set apart the microbeam analysis of high radioactive materials from standard practice.

Minor actinide transmutation in fast reactor metal fuels irradiated for 120 and 360 equivalent full-power days

An irradiation experiment on uranium–plutonium–zirconium (U–Pu–Zr) alloys containing 5 wt% or less minor actinides (MAs) and rare earths was carried out in the Phénix fast reactor. The isotope compositions of the fuel alloys irradiated for 120 and 360 equivalent full-power days (EFPDs) were chemically analyzed by inductively coupled plasma–mass spectrometry after 3.3–5.3 years of cooling. The results of chemical analysis indicated that the discharged burnups of the fuel alloys irradiated for 120 and 360 EFPDs were 2.1–2.5 and 5.3–6.4 at%, respectively. The changes in the isotopic abundances of plutonium, americium, and curium during the irradiation experiment were assessed to discuss the transmutation performance of MA nuclides added to U–Pu–Zr alloy fuel. Multigroup three-dimensional diffusion and burnup calculations accurately predicted the changes in these isotopic abundances after fuel fabrication. An evaluation of the MA transmutation ratio based on the results of chemical analysis revealed that the quantity of MA elements in the U–19Pu–10Zr–5MA (wt%) alloy decreased by about 20% during the irradiation experiment for 360 EFPDs.

Investigation of high temperature irradiated fuel-liquefied Zircaloy interactions in support of severe accident safety studies

The problem of irradiated fuel (both UO2 & Mixed Oxide Fuels) interactions with liquefied Zircaloy at high temperatures is central to the understanding of bundle degradation mechanisms during reactor power transients or severe accidents. These initial interactions of the cladding and the irradiated fuel result in a melt (corium) and then to a loss of bundle geometry and the corium accumulation in a pool. ITU investigated the interaction of irradiated fuel and compared it with non-irradiated fuel with its Zircaloy cladding at 2000 °C for various short times. This was its contribution to the COLOSS (Core Loss of Geometry) project carried out under an EC framework programme. The tests were investigated by optical microscopy with image analysis and then by SEM-EDS analysis. The dissolution of the irradiated fuel by the Zircaloy melt was very variable and heterogeneous, but for non-irradiated fuel was reasonably uniform and constant. The kinetics of the non-irradiated UO2-liquefied Zircaloy interactions was shown in another work package of the project to follow diffusion-limited mechanisms that could be modelled. The large variation in the results with the irradiated fuel rods made it difficult to model these interactions, nevertheless, they appear to have similar parabolic kinetics seen in non-irradiated fuel. The cracked condition of the fuel and the fission gas release during these interactions are major factors for fuel break-up, dispersion and dissolution in the melt under temperature transients.