Microstructural characterization of Irradiated U-Pu-Zr Fuels (original) (raw)
Related papers
Journal of Nuclear Materials, 2014
Microstructural changes in a set of commercial grade UO 2 fuel samples have been investigated using synchrotron based micro-focused X-ray fluorescence (l-XRF) and X-ray diffraction (l-XRD) techniques. The results are associated with conventional UO 2 materials and relatively larger grain chromia-doped UO 2 fuels, irradiated in a commercial light water reactor plant (average burn-up: 40 MW d kg À1). The lattice parameters of UO 2 in fresh and irradiated specimens have been measured and compared with theoretical predictions. In the pristine state, the doped fuel has a somewhat smaller lattice parameter than the standard UO 2 as a result of chromia doping. Increase in micro-strain and lattice parameter in irradiated materials is highlighted. All irradiated samples behave in a similar manner with UO 2 lattice expansion occurring upon irradiation, where any Cr induced effect seems insignificant and accumulated lattice defects prevail. Elastic strain energy densities in the irradiated fuels are also evaluated based on the UO 2 crystal lattice strain and non-uniform strain. The l-XRD patterns further allow the evaluation of the crystalline domain size and sub-grain formation at different locations of the irradiated UO 2 pellets.
Non-Destructive Characterization of UO2+x Nuclear Fuels
Microscopy Today, 2017
This article describes the effect of fabrication conditions on as-sintered microstructures of various stoichiometric ratios of uranium dioxide, UO 2 + x , with the aim of enhancing the understanding of the fabrication process and developing and validating a predictive microstructure-based model for fuel performance. We demonstrate the ability of novel, non-destructive methods such as near-field highenergy X-ray diffraction microscopy (nf-HEDM) and micro-computed tomography (μ-CT) to probe bulk samples of high-Z materials by non-destructively characterizing three samples: UO 2.00 , UO 2.11 , and UO 2.16 , which were sintered at 1450°C for 4 hours. The measured 3D microstructures revealed that grain size and porosity were influenced by deviation from stoichiometry.
Thermal Recovery of Radiation Defects and Microstructural Change in Irradiated UO2Fuels
Journal of Nuclear Science and Technology, 1993
Thermal recovery of radiation defects and microstructural change in U0 2 fuels irradiated under LWR conditions (burnup: 25 and 44 GWd/t) have been studied after annealing at temperature range of 450-l,SOO'C by X-ray diffractometry and transmission electron microscopy (TEM). The lattice parameter of as-irradiated fuels increase with higher burn up, which was mainly due to the accumulation of fission induced point defects. The lattice parameter for both fuels began to recover around 450-650'C with one stage and was almost completely recovered by annealing at 850'C for 5 h. Based on the recovery of broadening of X-ray reflections and TEM observations, defect clusters of dislocations and small intragranular bubbles began to recover around 1,150-1.450'C. Complete recovery of the defect clusters, however, was not found even after annealing at 1,800'C for 5 h. The effect of irradiation temperature on microstructural change of sub-grain structure in high burnup fuels was assessed from the experimental results.
High-energy synchrotron study of in-pile-irradiated U–Mo fuels
Scripta Materialia, 2016
Here synchrotron scattering analysis results on U-7wt%Mo fuel specimens irradiated in the Advanced Test Reactor to three burnup levels (3.0, 5.2, and 6.3×10 21 fission/cm 3) are reported. Mature fission gas bubble superlattice was observed to form at intermediate burnup. The superlattice constant was determined to be 11.7 and 12.0 nm by wide-angle and small-angle scattering respectively. Grain subdivision takes place throughout the irradiation and causes the collapse of the superlattice at high burnup. The bubble superlattice expands the U-Mo lattice and acts as strong sink for radiation-induced defects. The evolution of dislocation loops was, therefore, suppressed until the bubble superlattice collapsed.
Micro-structural study and Rietveld analysis of fast reactor fuels:U-Mo fuels
UeMo alloys in as-cast as well as in annealed conditions have been studied using Optical Microscope, SEM, XRD. The monoclinic a'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. The dendritic microstructure of g-(U,Mo) and B.C.C. 'Mo' phase of 33 at.% UeMo alloy have been analysed. Rietveld analysis has been done to optimize lattice parameters and calculate phase fractions in annealed alloys. The Vickers microhardness of U 2 Mo phase shows lower hardness than two phase microstructures in annealed alloys. a b s t r a c t UeMo alloys are the candidate fuels for both research reactors and fast breeder reactors. In-reactor performance of the fuel depends on the microstructural stability and thermal properties of the fuel. To improve the fuel performance, alloying elements viz. Zr, Mo, Nb, Ti and fissium are added in the fuel. The first reactor fuels are normally prepared by injection casting. The objective of this work is to compare microstructure, phase-fields and hardness of as-cast four different UeMo alloy (2, 5, 10 and 33 at.% Mo) fuels with the equilibrium microstructure of the alloys. Scanning electron microscope with energy dispersive spectrometer and optical microscope have been used to characterize the morphology of the as-cast and annealed alloys. The monoclinic a'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. A comparison of metallographic and Rietveld analysis of as-cast (dendritic microstructure) and annealed U-33 at.% Mo alloy, corresponding to intermetallic compound, has been reported here for the first time. This study will provide in depth understanding of microstructural and phase evolution of UeMo alloys as fast reactor fuel.
Nuclear Engineering and Design, 2015
An improved robust formulation for constituent distribution in metallic nuclear fuels is developed and implemented into the advanced fuel performance framework BISON. The coupled thermal diffusion equations are solved simultaneously to reanalyze the constituent redistribution in post irradiation data from fuel tests performed in Experimental Breeder Reactor II (EBR-II). Deficiencies observed in previously published formulation and numerical implementations are also improved. The present model corrects an inconsistency between the enthalpies of solution and the solubility limit curves of the phase diagram while also adding an artificial diffusion term when in the 2-phase regime that stabilizes the standard Galerkin Finite Element (FE) method used by BISON. An additional improvement is in the formulation of zirconium flux as it relates to the Soret term. With these new modifications, phase dependent diffusion coefficients are revaluated and compared with the previously recommended values. The model validation included testing against experimental data from fuel pins T179, DP16 and T459, irradiated in EBR-II. A series of viable material properties for U-Pu-Zr based materials was determined through a sensitivity study, which resulted in three cases with differing parameters that showed strong agreement with one set of experimental data, rod T179. Subsequently a full-scale simulation of T179 was performed to reduce uncertainties, particularly relating to the temperature boundary condition for the fuel. In addition a new thermal conductivity model combining all available data covering 0 to 100% zirconium concentration and a zirconium concentration dependent linear heat rate solution derived from Monte Carlo N-Particle (MCNP) simulations were developed. An iterative calibration process was applied to obtain optimized diffusion coefficients for U-Pu-Zr metallic fuels. Optimized diffusion coefficients suggest relative improvements in comparison to previous reported values. The most influential or uncertain phase is found to be the gamma phase, followed by alpha phase, and thirdly the beta phase; indicating separate effect testing should concentrate on these phases.
Journal of Nuclear Science and Technology, 1994
The effect of grain size on microstructural change and damage recovery in UO, fuels was studied by X-ray diffractometry (XRD) and transmission electron microscopy (TEM). The as-irradiated lattice parameter of the standard fuel (grain size : 16 pm) was larger than that of the large-grained fuel (43 pm), indicating a larger number of fission-induced point defects in the lattice of the former fuel. This tendency was in contrast to previously reported results for low burnup fuels below l G W d / t. The lattice dilation in the present high burnup fuels was mainly due to the accumulation of vacancies. The lattice parameter of both fuels began to recover around an irradiation temperature of 45O-65O0C, and both had a complete recovery a t 850°C. On annealing a t high temperatures of 1,450-1,8OO0C, the bubble diameter in the standard fuel was larger than that in the large-grained fuel. This indicated that vacancy diffusion from the grain boundaries plays an important role during bubble coarsening a t high temperatures.
Post irradiation examination of a uranium-zirconium hydride TRIGA fuel element
Frontiers in Energy Research
Low-enriched (LEU) U-ZrH fuel, with a235U content less than 20% of the total uranium, is being evaluated for possible use in different types of reactors, including space nuclear systems, light water reactors (LWRs) and micro-reactors. As a result, it is beneficial to better understand the macrostructural and microstructural changes that occur in this fuel during irradiation. This paper reports the results of the post irradiation examination of an LEU U-ZrH fuel element (30 wt.% U, <20% 235U) using neutron radiography, precision gamma scanning, chemical analysis, optical metallography and scanning electron microscopy combined with energy dispersive spectroscopy and wavelength dispersive spectroscopy, where the fuel element was irradiated in a Training, Research, Isotope, General Atomics (TRIGA) reactor. Results of microstructural characterization indicated some dehydriding and cracking of the U-ZrH fuel occurred during irradiation; an axial and radial burnup gradient existed in th...
Journal of Nuclear Materials, 2011
Low-enriched uranium-molybdenum (U-Mo) alloy particles dispersed in aluminum alloy (e.g., dispersion fuels) are being developed for application in research and test reactors. To achieve the best performance of these fuels during irradiation, optimization of the starting microstructure may be required by utilizing a heat treatment that results in the formation of uniform, Si-rich interaction layers between the U-Mo particles and Al-Si matrix. These layers behave in a stable manner under certain irradiation conditions. To identify the optimum heat treatment for producing these kinds of layers in a dispersion fuel plate, a systematic annealing study has been performed using actual dispersion fuel samples, which were fabricated at relatively low temperatures to limit the growth of any interaction layers in the samples prior to controlled heat treatment. These samples had different Al matrices with varying Si contents and were annealed between 450 and 525°C for up to 4 h. The samples were then characterized using scanning electron microscopy (SEM) to examine the thickness, composition, and uniformity of the interaction layers. Image analysis was performed to quantify various attributes of the dispersion fuel microstructures that related to the development of the interaction layers. The most uniform layers were observed to form in fuel samples that had an Al matrix with at least 4 wt.% Si and a heat treatment temperature of at least 475°C.
Journal of Nuclear Materials, 2009
Heavy ion irradiation has been proposed for discriminating UMo/Al specimens which are good candidates for research reactor fuels. Two UMo/Al dispersed fuels (U-7 wt%Mo/Al and U-10 wt%Mo/Al) have been irradiated with a 80 MeV 127 I beam up to an ion fluence of 2 Â 10 17 cm À2. Microscopy and mainly Xray diffraction using large and micrometer sized beams have enabled to characterize the grown interaction layer: UAl 3 appears to be the only produced crystallized phase. The presence of an amorphous additional phase can however not be excluded. These results are in good agreement with characterizations performed on in-pile irradiated fuels and encourage new studies with heavy ion irradiation.