Showing papers on "Zirconium alloy published in 2008"

Cladding embrittlement during postulated loss-of-coolant accidents.

Abstract: The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

Identification and characterization of a new zirconium hydride

Abstract: Zirconium alloys are currently used in nuclear power plants where they are susceptible to hydrogen pick-up. Hydride precipitation may occur when the hydrogen solubility limit is reached. Various Zr hydride phases, gamma, delta and epsilon have been identified since the 1950s. Combining electron precession microdiffraction, electron energy loss spectroscopy and ab initio electronic calculations, a new Zr hydride named zeta has been identified and characterized. It belongs to the trigonal crystal system with space group P3 m1 and it is fully coherent with the alphaZr matrix.

Characterization of Zirconium Hydrides and Phase Field Approach to a Mesoscopic-Scale Modeling of Their Precipitation

Abstract: Zirconium alloys are currently used in nuclear power plants where they are submitted to hydrogen pick-up. Hydrogen in solid solution or hydride precipitation can affect the behavior of zirconium alloys during service but also in long term storage and in accidental conditions. Numerical modeling at mesoscopic scale using a “phase field” approach has been launched to describe hydride precipitation and its consequences on the mechanical properties of zirconium alloys. To obtain realistic results, it should take into account an accurate kinetic, thermodynamic, and structural database in order to properly describe hydride nucleation, growth, and coalescence as well as hydride interaction with external stresses. Therefore, an accurate structural characterization was performed on Zircaloy-4 plates and it allowed us to identify a new zirconium hydride phase called ζ. The ζ phase has a trigonal symmetry and is fully coherent with hcp αZr. The consequences of this new zirconium hydride phase on hydride transformation process and stress-reorientation phenomenon are discussed. A first attempt to numerically model the precipitation of this new zirconium hydride phase has been undertaken using the phase field approach.

Hydrogen Content, Preoxidation, and Cooling Scenario Effects on Post-Quench Microstructure and Mechanical Properties of Zircaloy-4 and M5 ® Alloys in LOCA Conditions

Abstract: Previous papers pointed out the influence of long-term service exposures on the thermal-mechanical behavior of Zr alloys in LOCA conditions and, especially, the impact of in-service hydrogen pick-up on post-quench mechanical properties. Moreover, the oxide layer grown under in-service conditions was occasionally expected to have a protective effect against high temperature oxidation. Finally, the oxygen and hydrogen distributions within the prior-β layer appear as a key parameter with regard to the residual ductility of the alloy, especially as a function of the cooling scenario. The objective of the study presented here was to further investigate the influence of these parameters on the post-quench mechanical properties. Unirradiated Zircaloy-4 and M5® cladding tubes were consequently hydrided up to different concentration levels, then oxidized at high temperature (1000–1200°C) up to at least 10 % measured equivalent cladding reacted (ECR) and directly quenched to room temperature (RT). Ring compression tests (RCT), 3-point bending tests (3PBT) at RT and 135°C, as well as impact tests at RT were then performed to determine the evolution of the post-quench mechanical properties of Zircaloy-4 and M5® alloys with H content. Similarly, specimens preoxidized out-of-pile were also submitted to high temperature oxidation and direct quench, as well as to post-quench ring compression tests. Along with calculations of oxygen diffusion in the metal, results from those tests allowed us to estimate the assumed protective effect of the pretransient oxide layer. Finally, using specimens in the as-received condition or hydrided to typical end-of-life H contents, the effect of temperature history after oxidation at 1200°C was studied, i.e., at the end of the high temperature isothermal oxidation, samples were either submitted to direct quenching to RT or to slow cooling to different final quenching temperatures. It was thus demonstrated that the cooling scenario has a significant impact on the post-quench mechanical properties. All test samples were investigated by means of fractographic examinations to assess the type of failure mode. Moreover, a deep metallurgical analysis has been performed: SEM and image analysis were used for accurate phase thickness measurements, nuclear and electron microprobes for quantitative mapping of hydrogen and oxygen. It proved that the oxygen and hydrogen contents and their distribution in the prior-β layer have a first-order influence on the residual ductility. From all the results obtained on as-received and hydrided samples directly quenched from the oxidation temperature, it was then possible to derive a relationship between structural parameters, i.e., oxygen and hydrogen contents and thickness of the prior-β layer, and the post-quench impact properties at RT.

Microstructural Characterization of Oxides Formed on Model Zr Alloys Using Synchrotron Radiation

Abstract: 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, y) and control materials (pure Zr, Zircaloy-4) 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 Zr3O 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.

Kinetics of Hydrogen Absorption and Release in Zirconium Alloys During Steam Oxidation

Abstract: The kinetics of hydrogen absorption during steam oxidation in Zr–Sn and Zr–Nb alloys in the temperature range of 1,000–1,400 °C was investigated by neutron radiography. Hydrogen uptake can be subdivided into two steps: an initial phase and a state of equilibrium. The initial phase is controlled by the kinetics of hydrogen diffusion through the growing oxide layer. In the state of equilibrium, transport kinetics does not determine the hydrogen content of the material. An equilibrium is established between the hydrogen content of the gas environment and the metal phases. The temperature dependence of hydrogen absorption is Arrhenius-like at temperatures between 1,100 and 1,300 °C.

X-ray diffraction studies on asymmetrically broadened peaks of heavily deformed zirconium-based alloys

Abstract: The diffraction peaks of Zircaloy-2 and Zr–2.5% Nb alloys at various deformations are found to be asymmetric in nature. In order to characterize the microstructure from these asymmetric peaks of these deformed alloys, X-ray diffraction line profile analysis like Williamson–Hall technique, Variance method based on second- and fourth-order restricted moments and Stephens model based on anisotropic strain distribution have been adopted. The domain size and dislocation density have been evaluated as a function of deformation for both these alloys. These techniques are useful where the dislocation structure is highly inhomogeneous inside the matrix causing asymmetry in the line profile, particularly for deformed polycrystalline materials.

Electrochemical Conversion of Oxide Precursors to Consolidated Zr and Zr−2.5Nb Tubes

Abstract: Porous tubular oxide precursors have been fabricated from the ZrO 2 powder and its mixture with the Nb 2 O 5 powder, and directly metallised to pure Zr or the Zr-2.5Nb Zircaloy tubes through electrochemical reduction and deoxygenation that induces in situ consolidation or sintering of the metallized tubes in molten CaCl 2 at ∼900 °C. This new process is simple, fast, and low in energy consumption, promising a new technology for the fabrication of zirconium/Zircaloy tubes, which are the crucial materials in nuclear reactors and chemical plants. Also reported in this paper is the mechanism of the electrochemical process, correlating the cyclic voltammogram of ZrO 2 powder in a metallic cavity electrode with the morphological and compositional analyses of the products from potentiostatic electrolysis of porous ZrO 2 pellets.

Oxidation of Zirconium Alloys in Oxygen at High Temperatures up to 1600 °C

Abstract: The oxidation kinetics of the classic pressurized water reactors (PWR) cladding alloy Zircaloy-4 has been extensively investigated over a wide temperature range. In recent years, new cladding alloys optimized for longer operation and higher burn-up are being increasingly used in Western light water reactors (LWR). These alloys were naturally optimized regarding their corrosion behavior for operational conditions. The publicly available data on high temperature oxidation of the various cladding materials are very scarce. This paper presents the results of a first test series with Zircaloy-4 as reference material, Framatome Duplex cladding, Framatome M5® and the Russian E110 alloy. The first two are Zr–Sn, the latter two Zr–Nb alloys. All materials were investigated in isothermal and transient tests in a thermal balance under argon–oxygen atmosphere. Strong and varying differences (up to 500%) of oxidation kinetics between the alloys were found till 1000 °C, where the breakaway effect plays a role. Smaller but still significant differences (20–30%) were observed at higher temperatures. Generally, the advanced cladding alloys here studied show also a favorable behavior at high temperatures during accident scenarios.

Formation, mechanical properties and corrosion resistance of Ti-Pd base glassy alloys

Abstract: No biocompatible Ti-based glassy alloys without a harmful element have been reported. We have examined the mechanical and chemical properties of Ti–Pd–Zr–Si glassy alloy in comparison with pure Ti metal and Ti–6Al–4V alloy which have been used so far for biomaterials. The present Ti–Pd base glassy alloys do not contain Al and Ni elements which are considered to be rather toxic. Melt-spun Ti 45 Zr 50− x Pd x Si 5 glassy alloy ribbons ( x = 35, 40, 45) exhibited good bend ductility and had higher Vickers’s hardness and lower Young’s modulus as compared to pure titanium and Ti–6Al–4V alloy. In addition, the Ti 45 Zr 50− x Pd x Si 5 glassy alloys had higher corrosion resistance and were passivated over a wide range and at the lower passive current density of approximately 10 −2 Am −2 than at of pure titanium and Ti–6Al–4V alloy in 1 mass% lactic acid and PBS(−) solutions at 310 K.

Preparation and processing of platelet-reinforced ceramics by the directed reaction of zirconium with boron carbide

Abstract: Composites have been prepared by the directed reaction of a liquid metal with boron carbide to yield platelet-reinforced carbide matrix materials. Dense bodies with less than 1 percent porosity in sizes up to 8 in. in diameter and 1-2 in. thick have been prepared to near-net shape. The preparation and processing of these materials are discussed with specific examples in the zirconium system. The observed microstructures of these materials are also presented, together with a discussion of methods for tailoring the microstructure, and thus the properties, of the composite by varying the processing conditions. 9 refs.

Experimental and Modeling Approach of Irradiation Defects Recovery in Zirconium Alloys: Impact of an Applied Stress

Abstract: During neutron irradiation, both interstitial and vacancy loops are formed in high concentration in zirconium alloys Due to this high density of loops, the material is considerably hardened, but the recovery of the radiation damage during a heat treatment leads to a progressive softening of the irradiated material The recovery of the radiation induced hardening has been investigated using microhardness tests Transmission electron microscopy (TEM) observations performed on irradiated foils have also shown that the loop density falls while the loop size increases during the thermal annealing Furthermore, the TEM analysis has revealed that only vacancy loops are present in the material after long term annealing, the interstitial loops having entirely disappeared A numerical cluster dynamic modeling has also been used in order to reproduce the material recovery for various annealing conditions The microstructural evolution during mechanical testing with various loading conditions has also been studied It has been shown that during a creep test with low applied stress (130 MPa) and high temperature (450°C), the microstructure evolution can essentially be explained by the thermal recovery of the loops leading to glide of dislocations as found for an non-irradiated material At intermediate temperature (400°C), it is shown that for low stress level (130 MPa) the microstructure evolution can also be explained by the thermal recovery of loops, whereas for higher stress (250 MPa), sweeping of loops by gliding dislocations can also occur In addition, for an applied stress of 130 MPa and a temperature of 400°C, dislocation density is higher in the irradiated material than in the non-irradiated material deformed in the same conditions It is also shown that secondary slip systems are more activated in the irradiated material than in the non-irradiated material From this detailed analysis, the mechanical behavior during creep is interpreted in terms of microscopic deformation mechanisms

Synthesis and properties of bulk metallic glasses in the ternary Ni–Nb–Zr alloy system

Abstract: Bulk metallic glasses (BMGs) with high thermal stability, good mechanical properties and high corrosion resistance were synthesized in the Ni–Nb–Zr system. A large bulk glass-forming region with 60