High Temperature Oxidation of Zr-2.5%wt Nb Alloys Doped with Yttrium (original) (raw)
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Oxidation of Advanced Zirconium Cladding Alloys in Steam at Temperatures in the Range of 600–1200 °C
Oxidation of Metals, 2011
ABSTRACT The oxidation kinetics of the classical pressurized water reactors (PWR) cladding alloy Zircaloy-4 have been extensively investigated over a wide temperature range from operational conditions to beyond design basis accident (BDBA) temperatures. In recent years, new cladding alloys optimized for longer operation and higher burn-up are used in Western light water reactors (LWR). This paper presents the results of thermo-gravimetric tests with Zircaloy-4 as the reference material, Duplex DX-D4, M5® (both AREVA), ZIRLO™ (Westinghouse), and the Russian E110 alloy. All materials were investigated in isothermal and transient tests in a thermal balance with steam furnace. Post-test analyses were performed by light-microscopy and neutron radiography for investigation of the hydrogen absorbed by the metal. Strong and varying differences (up to 800%) in oxidation kinetics between the alloys were found at up to 1000°C, where the breakaway effect plays a role. Less but significant differences (ca. 30%) were observed at 1100 and 1200°C. Generally, the M5® alloy revealed the lowest oxidation rate over the temperature range investigated whereas the behavior of the other alloys was considerably dependent on temperature. A strong correlation was found between oxide scale structure and amount of absorbed hydrogen. KeywordsHigh-temperature oxidation–Zirconium alloys–Cladding–Light water reactor–Nuclear safety
Materials and Corrosion, 2013
This paper presents the results of parametric studies on the existing test procedures to assess nodular corrosion susceptibility of zirconium (Zr) base alloys. The parameters included the level of dissolved oxygen (DO) in the steam, exposure time, and type of exposure. The alloys studied were Zircaloy-2, Zircaloy-4, Zr-2.5Nb, and Zr-1Nb. A two-step test procedure involving prefilming at 410 8C followed by nodule growth at 510 8C in deaerated steam (using demineralized water) was found to be the most representative for Zircaloys and none of the test conditions produced nodules on Zr-Nb alloys. Presence of high dissolved oxygen was found to suppress nodule formation in Zircaloys. A detailed investigation on the morphology of individual nodules and nodule cross section is presented. Mechanism of nodular corrosion, in particular and localized oxidation of Zr alloys at 400 8C, in general, in oxidizing environments of nuclear reactors has been discussed in the light of the present results of nodular corrosion. Long term oxidation of Zircaloy-2 and Zircaloy-4 under low and high DO conditions are reported and compared with our earlier results on the effect of dissolved oxygen on the long term oxidation and hydrogen pickup behavior of Zr-Nb alloys.
Oxidation behavior of welded Zry-3, Zry-4, and Zr–1Nb tubes
Nuclear Materials and Energy
The Transient Reactor Test (TREAT) facility is a research reactor designed to simulate rapid transients to test new fuel designs. TREAT's cladding is exposed to unique conditions compared to normal water reactors. These conditions include: exposure to air at high temperatures (≥600°C), rapid heating (≈700°C/ s), and cladding geometry that includes chamfers and welds. This work investigates the effects of chamfering and welding on the oxidation behavior of zirconium alloys (Zircaloy-3, Zircaloy-4, and Zr-1Nb). Tube specimens were examined under isothermal and transient conditions in dry and humid air. The effect of weld type (tungsten inert gas or electron beam), the number of welds, and alloying elements are compared. Thermogravimetric analysis was used to collect mass gain data during isothermal oxidation and the data was used to quantify the oxidation rate constant and the activation energy of oxidation. Oxide behavior in the weld region, chamfered region, and bulk tube was measured and compared. The microstructure and secondary phase precipitates in EBW tubes before and after breakaway were characterized. The electron beam welded Zr-1Nb specimen was found to have the most favorable oxidation behavior under both isothermal and transient conditions. Zry-4 oxidized the most readily and was the most affected by mechanical deformation.
Oxidation behavior of Zirconium, Zircaloy-3, Zircaloy-4, Zr-1Nb, and Zr-2.5Nb in air and oxygen
Nuclear Materials and Energy
currently utilizes a legacy Zircaloy-3 cladding, which is no longer commercially available. TREAT is air cooled and routinely operates at temperatures well above that of traditional reactor designs. This study investigates the oxidation behavior of pure zirconium and its alloys (Zircaloy-3, Zircaloy-4, Zr-1Nb, Zr-2.5Nb) in Ar+20%O 2 and N 2 +20%O 2 atmospheres at temperatures ranging from 400-800°C to determine which alloy should be implemented as TREAT's cladding. While the oxidation behavior of zirconium based cladding materials has been extensively documented, this study focuses on direct comparison between legacy Zircaloy-3 and contemporary alloys using a flat plate geometry and similar conditions seen at the TREAT facility. In this work, thermogravimetric analysis was used to measure both steady state and breakaway oxidation, which was then used to calculate oxidation rate constants and activation energies of each material. Oxide thickness was evaluated through microscopy of oxidized specimen cross sections. The Zircaloy-3 and Zr-1Nb alloys were found to be the most resistant to oxidation under the conditions of this study, whereas the Zr-2.5Nb alloy was found to be the most susceptible.
Internal oxidation of Nb-Zr alloys over the range 1555-1768°C at low oxygen pressures
Le Journal de Physique IV, 1993
Three Nb alloys, containing 1 w/oZr, 2.5 w/oZr, and IOW-2.5Zr, were internally oxidized in the range of 1555 to 1768 O C at oxygen pressures ranging from 5 x to 1 x torr. Linear kinetics were measured suggesting that oxygen arrival at the surface and not oxygen diffusion in the substrate was rate controlling. Both tetragonal and monoclinic ZrOn formed, the tetragonal form being favored by high nucleation rates (lower temperatures), lower alloy content, and location in the reaction zone (small particles near the surface). Monoclinic ZrOp formed at higher temperatures and deeper within the reaction zone where larger precipitates formed. The high solubility product of Zr02 in Nb-Zr alloys gives rise to non-Wagnerian behavior so that the solute is not precipitated out at the reaction front, additional precipitation occurring after the reaction front has passed. This causes a variation in the precipitate volume fraction with distance in the zone. Experimental observations are discussed in terms of various models for non-classical internal oxidation.
Zirconium in the Nuclear Industry: 18th International Symposium, 2018
We have used a range of advanced microscopy techniques to study the microstructure, the nanoscale chemistry and the porosity in a range of zirconium alloys at different stages of oxidation. Samples from both autoclave and in-reactor conditions were available to compare, including ZIRLO TM , Zr-1.0Nb and Zr-2.5Nb samples with different heat-treatments. (Scanning) Transmission Electron Microscopy ((S)TEM), Transmission Kikuchi Diffraction (TKD) 1 and automated crystal orientation mapping with TEM 2,3 were used to study the grain structure and phase distribution. Significant differences in grain morphology were observed between samples oxidised in the autoclave and in-reactor samples, with shorter, less well-aligned monoclinic grains and more tetragonal grains seen in the neutron irradiated samples. A combination of Energy Dispersion X-ray (EDX) mapping in STEM and Atom Probe Tomography (APT) analysis of SPPs can reveal the main and the minor element distributions respectively. Neutron irradiation seems to have little effect on promoting fast oxidation or dissolution of β-Nb precipitates, but encourages dissolution of Fe from Laves phase precipitates. Electron Energy Loss Spectroscopy (EELS) analysis of the oxidation state of Nb in β-Nb SPPs in the oxide reveal the fully oxidised Nb 5+ state in the SPPs deep into the oxide, but Nb 2+ in the crystalline SPPs near the metaloxide interface. EELS analysis and automated crystal orientation mapping with TEM have also revealed Widmanstatten-type suboxide layers in some samples with the hexagonal ZrO structure predicted by ab initio modelling 4. The combined thickness of the ZrO suboxide and oxygen-saturated layers at the metal-oxide interface correlates well to the estimated instantaneous oxidation rate, suggesting that the presence of this oxygen rich zone is part of the protective oxide that is rate limiting in the key in the transport processes involved in oxidation 5. Porosity in the oxide has a major influence on the overall rate of oxidation, and there is much more porosity in the rapidly oxidising annealed Zr-1.0Nb alloy than found in either the recrystallised alloy or the similar alloy exposed to neutron irradiation.
Properties of zirconium alloys and their applications in light water reactors (LWRs)
Materials Ageing and Degradation in Light Water Reactors, 2013
This chapter highlights the various uses and properties of zirconium alloy cladding and structural components used in nuclear power light water reactors. Specifi c attributes including dimensional stability, corrosion resistance, irradiation effects and mechanical properties are discussed in detail.
Corrosion behavior of Zr alloys with a high Nb content
Journal of Nuclear Materials, 2005
The corrosion behavior of the Zr alloy with a high Nb content was evaluated in the water loop system containing 2.2 wppm Li and 650 wppm B. The characteristics of the precipitates were analyzed by transmission electron microscopy (TEM) and the oxide was characterized by an X-ray diffraction method using a synchrotron radiation source. On the basis of the results obtained by these measurements, the relationship among the oxidation behavior, the precipitate characteristics and the oxide properties was discussed. It was shown that the Cu addition was of benefit to the corrosion resistance of the Zr alloy with a high Nb content and the corrosion resistance of the Cu-containing alloy (Zr-1.5Nb-0.5Sn-0.2Fe-0.1Cu) was superior to that of the Cr-containing alloy (Zr-1.5Nb-0.5Sn-0.2Fe-0.1Cr). The fine b-Nb precipitates were found more frequently in the Cu-containing alloy than the Cr-containing alloy when heat-treated in the same condition. The fraction of the tetragonal zirconia in the region of the metal/oxide interface was higher in the Cu-containing alloy than the Cr-containing alloy, suggesting that the stabilization of the tetragonal phase in the oxide was promoted more when the smaller precipitates are incorporated into the oxide. It is concluded that the fine distribution of b-Nb is desirable for stabilizing the tetragonal phase in the oxide, thereby increasing the corrosion resistance of the Zr alloy with a high Nb content.
MICROSTRUCTURAL CHANGES IN OXIDIZED Zr-2.5%Nb ALLOY BY THERMAL TRANSIENTS
2009
The experimental simulation of LOCA scenarios in CANDU reactors assumes transients of temperature on the pressure tubes from fuel channels. In case of a postulated accident of this kind, which is supposed to occur after some hot years in normal operating conditions, the safety analysis should take into account the oxidised state of pressure tube and the changes of its microstructure during the ramps temperature as well. This paper investigates the micro-structural changes in oxide layers and material base (Zr-2.5%Nb) resulted from specific thermal transients. This study was realised on Zr-2.5%Nb alloy samples, which were previously isothermally oxidized at temperature of 700 o C for different time intervals. Afterwards, the resulted samples with variables thickness oxide layers were subjected to the various temperature transients at different heating/cooling rates. The oxidation process in steam was carried out in a thermobalance facility. Using the thermo gravimetric analysis (TGA)...