CAPRIX1 experiment was the first irradiation in PHENIX Fast Breeder Reactor (FBR) of a CAPRA type fuel (<< Consommation Accrue de Plutonium dans les RApides >> -Increased consumption of plutonium in fast reactors) with the aim to study the behavior under irradiation of a fuel with a high Pu content (45 %) as well as that of its cladding made of Optimized 15-15Ti austenitic steel AIM1. This paper describes the results of post-irradiation examinations performed on a CAPRIX pin by using the analytical capabilities developed at LECA-STAR hot cell facility. CAPRIX pin global behavior was good in comparison to standard low Pu-content pins without any limiting performance sign nor safety alteration. Annular pellets kept a good integrity despite, at peak power node a diameter increase and deformation of the central hole and an offset from the geometric center. The formation of columnar grains in the central area and the fission gas release were low due to the low linear heat rates and the presence of the large central hole. This study contributed to improving the knowledge on the behavior of mixed oxide FBR fuels with a very high Plutonium content and provided support for the modelling and development of fuel performance codes to enable the extension of their validation domains for this type of fuel.

Focussed ion beam - scanning electron microscope (FIB-SEM) 3D examination was conducted on three standard UO2 and one Cr doped UO2 high burn-up pressurized water reactor (PWR) fuel samples. This work complemented other microanalysis examination, including an electron backscattered diffraction (EBSD) work on the polished surface. A parallel article giving the EBSD results was submitted simultaneously. Together, they found, in all the central area of these high burn-up samples: (i) a restructuring of the initial grains into smaller sub-grains forming low angle boundaries and with crystal orientations around that of their parent grains; and (ii) intragranular bubbles mostly situated on these low angle boundaries. The FIB-SEM 3D examination showed how such inter-sub-grain bubbles start as small compact but also small lenticular bubbles, similar to typical small intergranular lenticular bubbles. With increasing burn-up, these lenticular bubbles get thicker and locally interlink to form more complex bubbles. However, no long distance networks, between the sub-grains or between the original grains, were found. Such networks could have been a path for part of the fission gases to reach the grain boundaries, the grain edges (the intersection line of three grain boundaries), and the rod free volumes. These FIB-SEM 3D examinations brought details on the intragranular and intergranular bubbles situation for each studied volume. The distribution of the intragranular bubbles according to their sizes and shapes was exposed. The central restructuring, studied in this work, is likely to play a role in the increase of the fission gas release fractions at high burn-up. This work is an incentive to study further this restructuring and the bubbles formed, combining different approaches.

This paper discusses the use of electron backscattered diffraction to characterize restructuring in a set of UO2 samples, irradiated in a pressurized water reactor at a burn-up between 35 and 73 GWd/tU, including standard UO2 samples and Cr-doped UO2 samples, to provide a better understanding of restructuring occurring both on the periphery and in the center of high-burn-up pellets. The formation of a high burn-up structure on the periphery of high burn-up UO2 was confirmed in our experiment. We found restructuring associated with bubble formation of all the samples in the central area, with higher irradiation temperatures when the burn-up exceeded 61 GWd/tU, regardless of their initial microstructure. This restructuring tended to progress with the increasing burn-up and to sub-divide the initial grains into sub-grains, with orientations close to that of the parent grains. Radial changes and differences between these samples showed that the burn-up and the temperature were not the only relevant parameters involved in restructuring.

Within the frame of the EMPIrE test, four monolithic mini-plates were irradiated in the ATR reactor. In two of them, the monolithic U(Mo) foil had been PVD-coated with Zr before the plate manufacturing. Extensive microstructural characterizations were performed on a fresh archive mini-plate, using Optical Microscopy (OM), Scanning Electron Microscopy (SEM) combined with Energy Dispersive Spectroscopy (EDS), Electron Backscattered Diffraction (EBSD) and Focused Ion Beam (FIB)/Transmission Electron Microscopy (TEM) with nano EDS. A particular attention was paid to the examination of the U(Mo) foil, the PVD coating, the cladding/Zr and Zr/U(Mo) interfaces. The Zr coating has a thickness around 15 μm. It has a columnar microstructure and appears dense. The cohesion of the cladding/Zr and Zr/U(Mo) interfaces seems to be satisfactory. An almost continuous layer with a thickness of the order of 100–300 nm is present at the cladding/Zr interface and corresponds to an oxidized part of the Zr coating. At the Zr/U(Mo) interface, a thin discontinuous layer is observed. It could correspond to locally oxidized U(Mo). This work provides a basis for interpreting the results of characterizations on EMPIrE irradiated plates.

In oxide nuclear fuels, at high burn-up or during high temperature periods such as ramp tests, out-ofpile heating tests, or any irradiations at high linear heat rates, fission gases can form micrometric or quasi-micrometric bubbles. During nominal operations, these bubbles participate to the pellet swelling, to the decrease of the fuel thermal conductivity and are involved in the mechanisms leading to fission gas release. During events involving a temperature increase, the resulting increase in the internal pressure of the bubbles might play a role in fuel fragmentation and in the opening of grain boundaries. The gas densities inside these bubbles are therefore one of the useful experimental information for the understanding of the fuel behaviour, and for the fuel behaviour code progress and validation. Two methods were developed to evaluate the gas density in the quasi-micrometric bubbles, using electron probe micro analyser, secondary ion mass spectrometry and focused ion beam scanning electron microscope together. The first method provides a mean gas density for all quasi-micrometric bubbles in a given area. The second method provides a gas density in a single selected bubble. In addition to the gas density, the 3D size and shape of the selected bubble is measured and can be related to the gas density result. In this work, these methods were applied to the bubbles formed in the centre of a PWR Cr doped UO2 at 38.8 GWd/tU after a ramp test in the Osiris reactor, with a 12 h plateau at 470 W/cm, and to the bubbles formed in a PWR Cr doped UO2 at 62.8 GWd/t(U) in the centre of the pellet and on the bubbles of the high burn-up structure on the rim. Both show the high pressures reached in these bubbles. (C) 2020 Elsevier B.V. All rights reserved.

Micro-cantilevers were prepared with a dual beam microscope SEM/FIB on different fresh and irradiated Pressurized Water Reactor (PWR) UO2 fuels. Their local fracture properties (fracture toughness and fracture stress) and Young's moduli were determined by bending tests, carried out with a nano-indenter in situ of a SEM chamber. Different microstructural features were characterized by testing three crystalline planes into single grains and grain boundaries. Inside grains, irradiation has a limited effect on the measured properties, whereas irradiation leads to a clear drop of the grain boundary resistance, related to the presence of irradiation defects.

For the first time, fracture properties of a UO2 fuel preliminary irradiated in a Pressurized Water Reactor (PWR) were characterized by bending tests. Measurements were made at room temperature and at a local scale on micro-cantilevers. Notched and un-notched specimens were prepared with a Focalized Ion Beam (FIB) microscope. Some cantilevers were milled into single grains of the polycrystalline UO2 fuel sample with specific crystallographic orientations chosen with Electron BackScatter Diffraction (EBSD) measurements, others were milled in order to test grain boundaries. Then, every specimen was tested up to failure with a nano-indenter in situ of the SEM/FIB chamber. The resulting brittle fracture parameters (fracture toughness and fracture stress) are compared and discussed. Grain boundaries exhibit a significantly lower fracture resistance than grains, and {111} crystallographic planes appear weaker than {100} and {110} planes in single grains.

Local fracture toughness of a polycristalline ceramic was characterized with the micro-cantilever bending method. Beams were milled in single grains of cubic zirconia using a Focused Ion Beam (FIB) microscope. Grains with specific crystallographic orientations were chosen using Electron Backscatter Diffraction (EBSD) measurements to test specific plane families. Notched micro-cantilever beams were loaded up to fracture using a nano-indenter. For fracture toughness evaluation, three different methods were considered: an analytical solution, an isotropic Finite Element Method (FEM) calculation, and an anisotropic FEM model. These three methods gave similar toughness values. A good agreement was found with literature data measured on cubic zirconia single crystals at a macroscopic scale. In this work, no significant difference was noticed for fracture toughness between {100}, {110} and {111} crystallographic plane families. Such method could be used to characterize the local fracture properties of a sample showing extensive cracking such as irradiated nuclear fuel.

The determination of local burn-up on irradiated fuel is a key parameter, which is directly connected to its neutronic behavior. In high burn-up UO2 fuel, the burn-up distribution is not homogeneous, the pellet edge having the highest burn-up. In order to measure fission products balance (like Cs, Nd and Nd) several methods are available in hot labs facilities, like spectroscopy or mass spectroscopy isotopic analysis on dissolved fuel sample. Burnup values are then evaluated from neutronic codes calculations. Unfortunately, these determinations are not able to characterize any radial evolution of burnup. EPMA (Electron Probe Micro Analysis) and SIMS (Secondary Ion Mass Spectrometry) are more suitable to carried out such evaluations. Using a shielded SIMS (CAMECA IMS6F), quantitative isotopic ratio distribution ( U Nd Nd 238 146 145 + ) is determined thanks to standard sample for which isotopic analyses are available after dissolution. After to calculations performed with the CEA APOLLO2 code, SIMS Burn-up profiles are compared to those obtained from EPMA Nd w% profiles along the same radius. Radial burn up evaluations are validated on low (38 GWd/tM) and high (72 GWd/tM) burn-up PWR fuels.

Among the microanalytical techniques, electron probe microanalysis (EPMA) is one of the most powerful. Its performances can be used to provide an accurate characterization. In the present article the differences between the EPMA of highly irradiated materials and standard EPMA are highlighted. It focuses on the shielded EPMA specificities. Then, the article presents the difficulties encountered during the sample preparation and the analysis (mainly due to the radioactive background). In spite of these difficulties, some valuable results can be provided by a shielded EPMA on the in-pile behavior of nuclear irradiated fuel. Some results of specific examples analyzed by EPMA in nuclear fuel research are presented.

A new method for the quantitative determination of the total xenon concentration in irradiated nuclear fuel is presented. The SIMS measurement of xenon enables the detection of the gas filling bubbles which are not detected with EPMA. The quantification is achieved using the EPMA data as reference at position where no or nearly no bubbles are detected. A new approach using the complementary information given by EPMA, SEM and SIMS is proposed, it opens new horizons for the characterisation of fission gases in irradiated nuclear fuel.

Within the context of reprocessing irradiated nuclear fuel from nuclear power reactors, fuel rods are cut into pieces for dissolution in fuel reprocessing plants so as to separate the fuel from the cladding material. The structural waste coming from cladding – called “hulls” – is conditioned in specifically-designed matrices for disposal. Standard characterisations of these hulls, such as neutron emission measurements or gamma spectrometry analyses, have already been performed on such samples. This paper describes a series of local characterisations that were performed on these hulls using various microanalysis techniques. Three techniques were used: a scanning electron microscope equipped with a wavelength dispersive X-ray detector, an electron probe microanalyser, and a secondary ion mass spectrometer. This paper describes the analytical methodology used to better understand the physico-chemical behaviour of the different radioelements contaminating the hulls.

The behaviour of gases produced by fission is of great importance for nuclear fuel operation. Within this context, an experimental method for the characterisation of the fission gas including gas bubbles in an irradiated UO2 nuclear fuel was developed in our laboratory using SIMS, EPMA and SEM results. SIMS and EPMA have been used to measure the radial distribution of xenon and SEM gives information on bubble formation across the fuel pellet radius. Using SIMS, xenon concentration can be determined in the matrix and in the bubbles. A quantification method, allowing the determination of the total inventory of xenon, is proposed and qualified with EPMA results. It is concluded that the complementary micro-analytical techniques SIMS, EPMA and SEM are very powerful tools for the characterisation of the fission gas bubbles in irradiated nuclear fuel.

. The formation of gas bubbles is a key issue for nuclear fuel behaviour. The first measurements of xenon during a depth profile in irradiated nuclear fuel evidenced some peaks over a non-zero baseline. In comparison with EPMA mappings of xenon and SEM image of the sputtered surface, it is proven that the peaks correspond to bubbles while the baseline corresponds to xenon included in the UO 2 matrix.