Synchrotron X-ray diffraction investigations on strains in the oxide layer of an irradiated Zircaloy fuel cladding (original) (raw)
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Zirconium in the Nuclear Industry: 18th International Symposium, 2018
The corrosion process (oxidation and hydriding) of the zirconium alloy cladding is one of the limiting factors on the fuel rod lifetime, in particular for the Zircaloy-4 alloy. The corrosion rate of this alloy shows indeed a great acceleration at high burn-up in Light Water Reactors. Understanding the corrosion behavior under irradiation for this alloy is an important technological issue for the safety and efficiency of LWRs. In particular, understanding the effect of irradiation on the metal and the oxide layers is a key parameter in the study of corrosion behavior of zirconium alloys. Zircaloy-4 samples have undergone helium and proton ion-irradiation up to 0.3 dpa forming a uniform defect distribution up to 1 µm deep. Both as-received and pre-corroded samples have been irradiated in order to compare the effect of metal irradiation to that of oxide layer irradiation. After irradiation, samples have been corroded in order to study the impact of irradiation defects in the metal and in preexisting oxide layers on the formation of new oxide layers. Synchrotron X-ray micro-diffraction and micro-fluorescence are used to follow the evolution of oxide crystallographic phases, texture and stoichiometry both in the metal and in the oxide in cross-section. In particular, the tetragonal oxide phase fraction, which has been known to have an important role on corrosion behavior, is mapped in both unirradiated and irradiated metal at the sub-micron scale and appears to be significantly affected by irradiation. These observations, complemented with electron microscopy analyses on samples in carefully chosen areas of interest are combined in order to fully characterize changes due to irradiation in metal and oxide phases of both alloys.
Proceedings of 18th …, 2005
Several high-burnup PWR-fuel-claddings associated to high corrosion levels have been subjected to RIA transients in the CABRI facility. The zirconia layer surrounding the Zircaloy-4 fuel rods exhibited a complex behavior during the RIA transients. Regularly spaced incipient cracks are initiated in the zirconia layer and stop at the Zircaloy interface beneath the oxide. These cracks are mainly initiated in the direction normal to the maximum principal stress. The crack density in post-test metallographies appears to be tightly linked to the maximum applied strain. This cracking process is sometimes followed by oxide spalling. Oxide spalling has a major impact on thermal-mechanical fuel rod behavior during such transients, thus the understanding and modeling of spallation is an issue that deserves attention. The present paper provides a quantitative analysis of the related data generated within the scope of the CABRI REP-Na tests, in the last decade, and more recently in the first tests of the CABRI Water loop program. A link with several other mechanical testing programs such as tensile tests, creep tests is also established.
Materials and Design, 2023
Chromium (Cr)-coated Zircaloy fuel cladding has been considered a promising candidate materials system for accident tolerant fuels. In this work, two types of Cr coatings produced by cold sprayed (CS) and physical vapour deposited (PVD) methods were studied. In particular, a novel combination of C-ring compression tests at room temperature (RT) and 345 • C in an inert gas environment and real-time X-ray micro-computed tomography (XCT) imaging was adopted to investigate the failure processes. Before testing, the crystal structure and local properties were fully characterized; post testing, ex situ scanning electron microscope (SEM) imaging were conducted to complement the XCT measurements in crack density. It was found that the failure processes in both coatings vary with temperature, as discussed in detail. The hoop strength of first coating cracks' formation of CS materials were higher than the PVD materials due to their higher interfacial roughness and distribution of splatted grains in CS coating. Based on a calculation of the first Dundurs' parameter from the measured local properties and observed crack arrest/deflection at coating/substrate interface, it was found that the cold sprayed coating-cladding material system has a higher interfacial toughness in terms of critical strain energy release rate due to its interlocking interfacial structure.
The assessment of the mechanical properties of the highly irradiated fuel claddings during an RIA (Reactivity Initiated Accident) has been carried out in the framework of the PROMETRA programme. Three main types of tests including burst tests, hoop and axial tensile tests, have been performed in CEA-Saclay hot laboratories in order to determine the cladding tensile properties used in the SCANAIR code. The representativeness of each test with regard to the RIA loading conditions can be addressed and analyzed in terms of strain or stress ratio. The present paper reports the high strain rate ductile mechanical properties of irradiated ZIRLO TM and M5 TM alloys derived from the PROMETRA program and their comparison to the stress-relieved irradiated Zircaloy-4. Results of specific analysis of the behaviour of the 6 cycle M5 TM and ZIRLO TM 75 GWd/tM for temperatures higher than 600°C are also presented.
Acta Materialia, 2017
Dislocation structures in neutron irradiated Zircaloy-2 fuel cladding and channel material have been characterized by means of high-resolution synchrotron x-ray diffraction combined with whole peak profile analysis and by transmission electron microscopy (TEM). The samples available for this characterization were taken from high burnup fuel assemblies and offer insight into the evolution of the dislocation structure after the formation of dislocation loops containing a c component. Absolute dislocation density values are about 4e15 times higher for the whole peak profile compared to TEM analysis. Most interestingly, the diffraction analysis suggests that the total dislocation density, as well as the a loop density, increases with fluence for the cladding material type. This trend is also inferred from a Williamson-Hall representation but contradicts the TEM observations. The c loop density evolution is more complicated and doesn't display any particular trend. In addition, the diffraction analysis highlights the presence of well-developed shoulders adjacent to the basal reflections and noticeable peak asymmetry particularly for the channel samples that experienced slightly lower operation temperatures than the clad. The findings are discussed in respect of the perceived irradiation induced growth mechanisms in Zr alloys.
Microstructural Characterization of Oxides Formed on Model Zr Alloys Using Synchrotron Radiation
Journal of ASTM International, 2008
To understand how alloy chemistry and microstructure impact corrosion performance, oxide layers formed at different stages of corrosion on various model zirconium alloys ͑Zr-xFe-yCr, Zr-xCu-yMo, for various x, y͒ and control materials ͑pure Zr, Zircaloy-4͒ were examined to determine their structure and the connection of such structure to corrosion kinetics and oxide stability. Microbeam synchrotron radiation diffraction and fluorescence of oxide cross sections were used to determine the oxide phases present, grain size, and orientation relationships as a function of distance from the oxide-metal interface. The results show a wide variation of corrosion behavior among the alloys, in terms of the pretransition corrosion kinetics and in terms of the oxide susceptibility to breakaway corrosion. The alloys that exhibited protective behavior at 500°C also were protective during 360°C corrosion testing. The Zr-0.4Fe-0.2Cr model ternary alloy showed protective behavior and stable oxide growth throughout the test. The results of the examination of the oxide layers with microbeam X-ray diffraction show clear differences in the structure of protective and nonprotective oxides both at the oxide-metal interface and in the bulk of the oxide layer. The nonprotective oxide interfaces show a smooth transition from metal to oxide with metal diffraction peaks disappearing as the monoclinic oxide peaks appear. In contrast, the protective oxides showed a complex structure near the oxide-metal interface, showing peaks from Zr 3 O suboxide and a highly oriented tetragonal oxide phase with specific orientation relationships with the monoclinic oxide and the base metal. The same interfacial structures are observed through their diffraction signals in protective oxide layers formed during both 360°C and 500°C corrosion testing. These diffraction peaks showed much higher intensities in the samples from 500°C testing. The results for the various model alloys are discussed to help elucidate the role of individual alloying elements in oxide formation and the influence of oxide microstructure on the corrosion mechanism.
Zirconium Alloys for Fuel Element Structures
CHIMIA, 2005
Today more than 400 light water power reactors (LWRs) operate worldwide providing approximately 17% of the world's electricity demand. One important component for their successful operation is the fuel tube, made out of a zirconium alloy. A huge number of out-of-pile and in-pile experiments have been performed to improve step by step the fuel for higher burn-up and to reduce the failure rates of fuel pins close to zero. The influencing parameters for excellent or poor cladding behaviour are numerous and sometimes counteract each other. The process of cladding corrosion is slow, difficult to follow, the mechanistic understanding at best incomplete. A vast amount of literature documents the abundant tests and comes up with hypotheses and models for the materials behaviour. PSI has supported for the past 20 years the development of high burn-up fuel cladding by microstructural research studies and service work in post-irradiation examination of test pins. This article reviews the d...
Using Small X-Ray Beams to Understand Corrosion in Nuclear Fuel Cladding
X-Rays in Mechanical Engineering Applications, 2004
Uniform oxidation by the primary circuit water and associated may soon limit the service of Zr alloy fuel cladding in Light Water reactors. Understanding the differences in corrosion rate between alloys based on the microstructure of the protective oxide may allow us to design better alloying materials for severe duty cycle applications. The use of synchrotron radiation microbeam at APS allows the study of these oxide layers with an unique combination of the wealth of diffraction and fluorescence information and the level of spatial resolution obtained. We will discuss some of our experimental results and the potentials of these techniques in solving engineering problems.
As part of studies conducted in France on Reactivity Initiated Accident (RIA), IRSN and EDF have launched a large experimental project (PROMETRA) carried out by CEA in order to provide both material properties and material failure data [1]. During the first phase of a RIA event, the in-service loading deforms the cladding in the circumferential direction under multiaxial tension, in a situation close to an axial plane strain situation. In order to accurately evaluate the risk of rod failure during this stage, it is important to develop models able to predict the material behaviour under those representative loading conditions. Obviously, the fracture behaviour has also to be determined. To this end, uniaxial tensile data have been obtained between 20°C and 1100°C under high strain rates (0.01 to 5s -1 ) and high heating rates (up to 200°C.s -1 ) from specimens machined along the axis of the cladding or in the circumferential direction (ring specimens).