Effect of Grain Size on Microstructural Change and Damage Recovery in UO2Fuels Irradiated to 23 GWd/t (original) (raw)
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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.
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.
Journal of Nuclear Science and Technology, 1993
Fission gas behavior of U0 2 fuel pellets with controlled microstructure irradiated to 23 GW d/t in a test reactor has been studied by using a postirradiation annealing experiment. Four types of fuel pellets with or without additives were examined: (1) un-doped standard (grain size: 1611m), (2) un-doped large grained (43 pm), (3) 0.7 wt?r;' Nb 20 5-doped large grained (110 pm), (4) 0.2 wt?.;' Ti0 2-doped large grained (85 pm) fuels. The annealing was conducted at 1,600 or 1,800oC for 5 h in reducing or oxidizing atmospheres. Fission gas release and bubble swelling caused by the high temperature annealing for the two un-doped fuels were reduced to about 1/3-1/2 by increasing the grain size from 16 to 4311m, which roughly corresponded to the ratio of their grain sizes. By contrast, the performance of the two large grained fuels doped with Nb20 5 or Ti02 was roughly equivalent to, or rather inferior to that of the standard fuel, despite their large grain sizes of 110 and 85 pm. The fission gas behavior of un-doped fuels was aggravated by increasing the oxygen potential in the annealing atmosphere, while that of additive doped fuels did not depend on it. The effects of grain size. additive doping and oxygen potential on fission gas release and bubble swelling were discussed in connection with the diffusivities of fission gas atoms and cation vacancies.
A Monte Carlo model of irradiation-induced recrystallization in polycrystalline UMo fuels
Journal of Nuclear Materials, 2019
Experiments show that irradiation-induced recrystallization speeds up the swelling kinetics in U10wt%Mo fuels. However, recrystallization mechanisms and the effect of initial grain microstructures on recrystallization kinetics are still unclear. In this work a Monte Carlo model coupling the rate theory of defect evolution has been developed to study the irradiation-induced recrystallization. The rate theory is used to describe the spatial evolution of gas bubbles, interstitials and interstitial loops; First-Passage Kinetic Monte Carlo (FPKMC) approach is used to describe the fast and strongly anisotropic migration of interstitials, and a Cellular Automata method is used to model the formation of recrystallized grains. With the assumption that recrystallization may occur when the local interstitial loop density is larger than a given critical value, simulation results reveal that 1) recrystallized grains first nucleate on grain boundaries and the recrystallization zone front moves to the center of original coarse grains in the UMo matrix, and 2) recrystallization starts earlier in coarse polycrystalline structures, while the overall recrystallization kinetics decreases with increasing grain size. These results agree with experimental observations. The comparison of recrystallization kinetics obtained from experiments and modeling suggests that the interstitial loop accumulation leads to the recrystallization and the interstitial loop growth is suppressed inside coarse grains due to the over-pressured intra-granular gas bubbles.
Effects of pellet microstructure on irradiation behavior of UO2 fuel
Journal of Nuclear Materials, 1997
In-reactor tests and post-irradiation examinations (PIEs) were performed for standard and large-grained pellets with and without additives being soluble in a matrix and/or precipitated in a grain boundary, to confirm the effects of large grain structure on decreasing fission gas release (FGR) and swelling and to evaluate the influence of the additives in the matrix/grain boundary on them. The standard and large-grained pellets were loaded into smail-diameter rods equipped with a pressure gauge. These rods were irradiated to about 60 GWd/t U at a linear heat rate of about 30-40 kW/m in the Halden reactor and then subjected to PIEs. Large-grained pellets showed a smaller FGR compared with standard pellets. Post-irradiation annealing tests suggested that swelling during transient power was decreased for large-grained pellets, except for those with additive enhancing cation diffusion.
Journal of Nuclear Materials, 2018
The dimensional changes of a nuclear fuel in operation are strongly determined by two opposite effects. One of them is due to contraction of the as-fabricated pores, giving place to densification which is evident during the first stages of irradiation. This effect is counteracted by the swelling phenomenon provoked by the fission products that progressively accumulate in the fuel material. A model to evaluate the changes in fuel pellets porosity due to radiation and thermal effects taking into account the point defects flow to and from the pores is presented. A simplification of the model to assess the progress of porosity in isothermal re-sintering tests is also given. Simulations are compared with experimental data measured on UO 2 fuel pellets with a variety of microstructures at different temperatures and radiation conditions, attaining a good agreement.
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.
The development of a mechanistic code on fission product behavior in the polycrystalline UO2 fuel
Nuclear Engineering and Design, 2000
The current status of a mechanistic code (RTOP) on fission product behavior in the polycrystalline UO 2 fuel is described. Outline of the code and implemented physical models is presented. The general approach to the code validation is discussed. It is exemplified by the results of validation of the models of oxidation and grain growth. The different models of intragranular and intergranular gas bubbles behavior have been tested and the sensitivity of the code in the framework of these models has been analyzed. An analysis of available models of the resolution of grain face bubbles is also presented.
A semi-empirical model for the formation and depletion of the high burnup structure in UO2
In the rim zone of UO2 nuclear fuel pellets, the combination of high burnup and low temperature drives a microstructural change, leading to the formation of the high burnup structure (HBS). In this work, we propose a semi-empirical model to describe the formation of the HBS, which embraces the polygonisation/recrystallization process and the depletion of intra-granular fission gas, describing them as inherently related. For this purpose, we performed grain-size measurements on samples at radial positions in which the restructuring was incomplete. Based on these new experimental data, we infer an exponential reduction of the average grain size with local effective burnup, paired with a simultaneous depletion of intra-granular fission gas driven by diffusion. The comparison with currently used models indicates the applicability of the herein developed model within integral fuel performance codes.