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Acoustic microscope schematic and operational modes: (a) schematic of an acoustic microscope; (b) acoustic waves focused on the surface of the specimen in the surface imaging mode; (c) acoustic waves focused below the surface of the sample in the sub-surface imaging mode; (d)the acoustic sensor approaches the sample and the acoustic waves are gradually defocused in the acoustic signature mode.
Source publication
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 samp...
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Citations
... R&D in nuclear forensic science now focuses on developing new approaches that can identify signatures from nuclear and other radioactive materials using fast, inexpensive, non-destructive, and easy-to-use technologies and analytical methods [3,6]. Techniques such as computer vision and image analysis allow rapid determination of morphological and intensity characteristics of objects or complex structures [16][17][18][19][20]. The possible correlation between morphological characteristics of different UOC powders and their processing or production history has recently been investigated as a new nuclear forensic signature. ...
... R&D in nuclear forensic science now focuses on developing new approaches that can identify signatures from nuclear and other radioactive materials using fast, inexpensive, non-destructive, and easy-to-use technologies and analytical methods [3,6]. Techniques such as computer vision and image analysis allow rapid determination of morphological and intensity characteristics of objects or complex structures [16][17][18][19][20]. The possible correlation between morphological characteristics of different UOC powders and their processing or production history has recently been investigated as a new nuclear forensic signature. ...
... Moreover, high burnup fuels are characterised by larger volumes due to the swelling caused by the porosity increase, together with the precipitation of insoluble solid fission products and noble fission gases. This in-reactor matrix volume expansion is about 0.8-1 % per 10 GWd t −1 HM [16][17][18] . The nuclear fuel swelling causes the pellet to enter in contact with the cladding originating pellet-cladding chemical and mechanical interactions (PCI): the fuel becomes progressively bonded to the cladding while oxygen migrates from the pellet to the Zircaloy and an oxide layer progressively forms. ...
... Concerning the properties of the UO 2 , the density of the rim, due to the high burnup structure [61,62] , is entered in the input file as 0.75 times the UO 2 theoretical density (10960 kg m -3 ). The correction factor is derived by inserting a fractional porosity (p) of 0.21 reported in [63] in the empirical correlation introduced in [18,64] : ...
Among the factors influencing the degradation of the spent nuclear fuel cladding in interim dry storage, the irradiation history and the average burnup at discharge must be considered. In fact, due to the pellet-cladding contact particularly affecting high burnup fuels, the inner surface of the cladding becomes increasingly exposed to the damage caused by the alpha decaying actinides present at the rim of the pellet. Moreover, due to the low temperature conditions characteristic of the interim dry storage, thermal recombination of the produced defects is not expected to occur. Here, we investigate the irradiation damage accumulated inside an irradiated Zircaloy-4 cladding 32, 55 and 100 years after the end of irradiation and discharge from the reactor core. The considered cladded UO2 pellet belongs to a Pressurised Water Reactor (PWR) fuel rod consisting of five segments and having an average burnup at discharge of 50.4 GWdtHM−1. The calculations performed with Fluka 2021.2.0 Monte Carlo code show that the volume mostly affected by the irradiation damage corresponds to the ZrO2 layer formed between the pellet and the cladding. The actinides which are responsible for the alpha damage are mainly 242Cm, 244Cm, 241Pu and 238Pu. The recoiling daughter nuclei during the alpha decays produce irradiation damage only within the first μm of oxide layer.
... Contrary to those made in zirconia and Zr, they do not compare well with the measurements of Terrani et al., despite the close indentation depths (around 200 nm): 8 GPa and 120 GPa, respectively. Our measurements are however in good agreement with previous measurements obtained by acoustic microscopy where a mean decrease of Young's modulus from about 220 GPa to 150 GPa was recorded for a burnup increase from 0 to 100 GWd/tU [42][43][44]. Differences with Terrani's works could be related to the location of the measurements: in this study, indentation tests were performed very close to the FCI (at not more than 20 µm from the Zr|ZrO2 interface), while Terrani et al. have performed measurements on the whole pellet radius from which the average values were reported [20]. ...
During irradiation in a nuclear reactor, the fuel swelling and the cladding creep are at the origin of the contact between the fuel and the cladding, which in turn leads to the initiation of Zr oxidation by the UO2 fuel. In high burnup fuels, the zirconia layer formed at the Fuel Cladding Interface (FCI) presents numerous circumvolutions and a heterogeneous microstructure within its thickness. In this paper, nanoindentation measurements are carried out in the fuel-cladding interface of a PWR fuel rod irradiated up to a high burnup of 61 GWd/tU. The measured hardnesses and Young's moduli of the internal zirconia layer are compared to reference measurements obtained on a non-irradiated sample of commercial Y-TZP. The study of the evolution of mechanical properties across the zirconia layer led to the identification of four distinct zones related to the zirconia microstructure (phase, grain size, porosity and fission product recoil). Close to the cladding, the zirconia hardness HZrO2tends toward the zirconium hardness HZr. Then, the zirconia hardness increases significantly with the grain size, until reaching a plateau at the middle of the zirconia layer, where the maximum hardness is obtained. Following the same trend, the elastic modulus progressively increases in the zirconia layer and reaches a plateau near the fuel High Burnup Structure.
... / for the solid swelling per 1% of burnup [1], and a variable amount as a function of burnup for the gaseous case, depending on the local temperature history [6]. At the end of fuel life at typically 5% burnup, the combination of the two effects gives a total decrease of the density of about 6% on average in the whole pellet [8]. The density is not homogeneous within the pellet, both due to local variation of the fission rate and due to the subsequent evolution of the fission gasses produced. ...
Nuclear fuel undergoes several thermo-mechanical changes during
irradiation in a nuclear reactor, such as change of density, caused by solid
and gaseous swelling. This affects the heat transport within the pellet and,
when leading to the pellet-cladding gap closure, it also affects the gap
conductance, causing stress in the cladding.
The density of irradiated fuel pellets can be measured in post-irradiation
examination using several methods. In this work, a feasibility study was made
using the gamma-ray transmission micro-densitometry technique. This is
based on the comparison of two intensity measurements, with and without a
sample with well-characterized thickness. Using a collimated source, a local
examination of the density can be performed, scanning a pellet slice radially.
The proposed technique aims to obtain a spatial resolution of cca. 100
microns.
In this work, the parameters of the setup, such as the source activity, detector
counting time, slit dimensions, collimator length, and sample thickness, are
used to predict detector efficiency and expected count rates. The obtainable
precision of the density is assessed by first-order uncertainty propagation of
counting errors in the gamma-ray detection to the density estimate.
A collimator design was presented that achieves a reasonable compromise
between time requirements, precision, and spatial resolution. The sensitivity
of the performance to set-up parameters was investigated. In addition, a
realistic setup was modelled in MCNP6 for validation of the peak count-rate,
and to ensure that the total spectrum count-rate is within typical throughput
capabilities of HPGe detectors. The MCNP model was also used to confirm
that the assumed attenuation law is valid in a relevant geometry, and to
assess the spatial resolution, using the 10-90% edge spread of an Edge
Spread Function.
It is concluded that fuel density can be determined with <1 % precision, using
a 100-micron wide slit, and 1 hour of measurement.
... The onset of recrystallization occurs along preexisting grain boundaries, after which it continues to the grain interior until the grain is consumed [28,29] . Fission gas atoms such as Xe and Kr precipitate into bubbles and have a high affinity for defect locations (i.e., pores formed from the production of voids) [30] . The size of fission gas bubbles is proportional to the absorption rate of fission gas atoms continually produced during fission reactions [30] . ...
... Fission gas atoms such as Xe and Kr precipitate into bubbles and have a high affinity for defect locations (i.e., pores formed from the production of voids) [30] . The size of fission gas bubbles is proportional to the absorption rate of fission gas atoms continually produced during fission reactions [30] . Swelling due to fission gas bubble nucleation is higher in recrystallized fuel [28] . ...
Irradiation conditions such as fission power, fission rates, and temperature are important parameters for nuclear fuels because they influence the microstructural behavior and ultimate irradiation performance. Two U-7Mo/Al-3.5 wt.% Si dispersion fuel plates were irradiated in the Advanced Test Reactor edge-on to the core in the RERTR-9B experiment at different powers to study the effects of power on fuel phase swelling behavior. The results of non-destructive measurements have been reported in a previous paper [1], and the destructive examination results are being reported here. In particular, the microstructural evolution of the U-Mo fuel phase in irradiated fuel plates has been investigated by cutting transverse cross-sections, using generated scanning electron microscopy micrographs, and image processing methods to assess the changes in microstructural features, such as fission gas pore growth and non-recrystallized grain fraction. In this study, the focus was to examine the evolution of fission gas pores from both the constrained and unconstrained regions of U-Mo fuels irradiated at different powers. Based on an estimated linear gradient, the rate of change of fission gas pore size with locally calculated fission densities is more prominent in the higher power specimen. As such, the average porosity and pore area is approximately 11% and 2 × larger in the fuel irradiated at higher power than that of the lower power specimen. The growth of fission gas pores can accelerate the pore interconnectivity and thus fission gas outside the fuel kernel (FGO), which are two precursors for swelling in U-Mo fuels. The influence of increased power, internal plate stresses, and fission-induced creep on fuel kernel deformation and lateral mass transfer is also highlighted.
... In addition to the changing elemental composition of nuclear fuel due to burnup, it is also well known that fuel swelling occurs (Marchetti et al., 2017), (Schrire et al., 1998), (Middleburgh et al., 2012). In accordance with Eq. (1), changes in the density, ρ, will also affect the transmission of gamma-rays through the fuel. ...
Nondestructive gamma-ray spectrometry of nuclear fuel is routinely performed in axial gamma scanning devices and more recently with gamma emission tomography. Following the irradiation of a fresh nuclear fuel with high intensity neutron flux in a nuclear reactor core, a great number of gamma-emitting radionuclides are created. These can be utilized for gamma spectrometric techniques. However, due to the high density and atomic number of the nuclear fuel, self-attenuation of gamma-rays is a challenge, which requires attenuation correction in order to perform accurate analysis of the source activity in the fuel.
In this study, the degradation of the gamma-ray mass attenuation with burnup was investigated and, in addition, a predictive model was created by investigating the attenuation change at various gamma energies caused by the burnup of the nuclear fuel. This model is intended for use by spectrometry practitioners inspecting nuclear fuel. To this aim, the energy-dependent gamma-ray mass-attenuation coefficients were investigated as a function of burnup for UOX, and three MOX fuels having different initial Pu contents. The Serpent 2 reactor physics code was used to simulate the burnup history of the fuel pins. The nuclide inventory of the Serpent 2 output is combined with the NIST XCOM database to calculate the mass attenuation coefficients.
The mass attenuation coefficient of the fuel was found to decrease with the fuel burnup, in the range of a few percent, depending on the burnup and gamma energy. Also, a theoretical burnup dependent swelling model was imposed on fuel density to see how linear attenuation coefficient of fuel material is changed. Furthermore, greater effect may be expected on the transmitted intensity, where a simulation study of a PWR assembly revealed that the contribution from the inner rods in a scanned fuel assembly increased by tens of percent compared to the one with non-irradiated fresh fuels, when shielded by the outer rods of the assembly. A sensitivity analysis was performed in order to test the effect of a number of geometrical and operational reactor parameters that were considered to potentially effect the mass attenuation coefficient. Finally, a simple-to-use predictive model was constructed providing the mass-attenuation coefficient [cm²/g] of fuel as a function of burnup [MWd/kgHM] and initial Pu content [wt%]. The resulting predictive model was optimized by using the nonlinear regression method.
... This point will be further studied in future works. 2) Mechanical properties: Acoustic signatures [5] were realized on the 2 silica elements. Rayleigh velocity and Young modulus estimations are presented in negligible maximum variation of 1% after irradiation. ...
The fuel element plates of research reactors are subjected to swelling phenomena. These structure modifications impact an initial inter-plate distance of 1.8 mm. An ultrasonic device has been developed to investigate this parameter. With a 1 mm thickness, it relies on two transducers linked to an
electronic system. The feasibility of the distance measurement has been proved in a previous study and the irradiation impact on the transducer components is here studied. To do so, a radiation resistance experiment was realized in the Arc-Nucleart Institute of Grenoble. It allowed the study of the influence of radiations on the device active and passive components’ characteristics.
... Values of 232 GPa and 0.3 were selected, based on the VHR approximation grounded on elastic properties calculations at the density functional theory level [39]. Note that actual elastic properties of irradiated UO 2 may differ from those of the pristine material but the change measured in the fuel at the end of one burning cycle was only ~10 % [40]. Hydrostatic and biaxial stresses are related through the following equation [35]: ...
UO2 polycrystals were irradiated in the nuclear energy-loss regime, Sn (900 keV I and 2 Mev Au ions) and also with an additional electronic energy deposition, Se (900 keV I and 36 MeV W ions simultaneously, i.e. Sn&Se). The strain/stress state exhibited by the irradiated pellets was determined by x-ray diffraction measurements. Results show that both measured strain and estimated stress are lower in the dual-beam irradiated samples, indicating that there is an ionization-induced change in the ballistically-generated-defect spectrum. Furthermore, it is shown that the thin irradiated layers maintain the lattice parameter of the pristine material in the basal plane.
... The pellets are modelled as cylindrical solid elements with a flexible stiffness behaviour [11]. Ceramics (pellets) tend to be weak in tension, but strong in compression [15]. Consequently, the ANSYS \cast-iron" material model has been used to describe the pellets constitutive law i. ...
This paper presents the methodology applied for the experimental and numerical investigation of the mechanical response of spent nuclear fuel rods under static and dynamic loads. The experimental activities were conducted at the JRC Karlsruhe where a 3-point bending test device and an impact tower have been developed and commissioned at the hot-cell facilities. Results are provided for two PWR samples. Load-displacement curves describe the mechanical response of the sample in the 3point bending tests, whereas an image analysis methodology has been developed to comprehend the sample’s behaviour under dynamic loads (recorded using a high-speed camera). Finite Element Analysis (FEA) are used to simulate the rod’s response based on static and transient structural models in ANSYS Mechanical.
