Sensitivity to chemical composition variations and heating/oxidation mode of the breakaway oxidation in M5® cladding steam oxidized at 1000°C (LOCA conditions) (original) (raw)
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During a Loss of Coolant Accident (LOCA), the high temperature oxidation kinetics and the residual Post-Quench (PQ) mechanical properties of zirconium alloy fuel claddings may be impacted by the oxide layer formed under in-service conditions. This paper deals with the influence of such a pre-transient oxide layer on the further single-side steam oxidation kinetics at 1000 and 1200°C and the PQ mechanical properties of Zircaloy-4 and M5™ cladding with various thicknesses of pre-oxide layers (ranging from 10 to 35 µm). Zircaloy-4 and M5™ cladding samples, double-side pre-oxidized in a typical pressurized water medium chemistry up to about 1400 days at 340, 350 or 360°C, were one-side steam oxidized at 1000 or 1200°C then directly quenched in water at room temperature. Ring compression tests were then performed at 135°C to evaluate PQ mechanical properties of the materials. Finally, metallurgical observations were carried out to measure the thicknesses of the oxide and α Zr (O) layers and the oxygen diffusion profiles. For short oxidation times at 1000 or 1200°C, weight gains obtained for pre-oxidized samples are lower than those of non pre-oxidized materials: the formation of a "fresh" oxide at 1000 or 1200°C at the metal/oxide interface is delayed in the presence of a pre-oxide. These lower weight gains are however not associated with better PQ mechanical properties: the overall quantity of oxygen atoms which diffuse into the sub-oxide metallic layers at high temperature does not depend on the source of oxygen (reduction of the pre-oxide or formation of a "fresh" oxide). The "breakaway" resistance of Zircaloy-4 and M5™ at 1000°C appears not to be significantly modified by pre-oxidation under simulated in-service conditions.
Oxidation of Metals, 2020
The steam oxidation behavior of Zr-based Zircaloy-4 fuel cladding was studied at 1273 K with two different surface roughness levels. Steam was introduced either at room temperature (RT) or at 1273 K. Weight gain kinetics were evaluated by post-test weight measurement, and the reaction products and alloy microstructure were evaluated using optical microscopy. Hydrogen pickup was measured by the gas extraction technique. Specimen surface roughness did not affect the oxidation kinetics or the hydrogen absorption. The time to breakaway oxidation was suppressed when steam was introduced at RT, and the oxide was more adherent, suggesting superior mechanical properties. When steam was introduced at 1273 K, an undulated oxide-metal interface formed earlier and a higher amount of hydrogen was absorbed by cladding before the kinetic transition. The alloy grain grew into larger size in the condition when steam was injected at 1273 K compared to the condition when steam was injected at RT, which may affect the observed behavior. After the oxide breakaway, the rate of hydrogen absorption accelerated substantially independent of the temperature of steam injection.
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
Journal of Nuclear Materials, 2009
The influence of the oxide layer morphology on the hydrogen uptake during steam oxidation of (Zr,Sn) and Zr-Nb nuclear fuel rod cladding alloys was investigated in isothermal separate-effect tests and large-scale fuel rod bundle simulation experiments. From both it can be concluded that the concentration of hydrogen in the remaining metal strongly depends on the existence of tangential cracks in the oxide layers formed by the tetragonal -monoclinic phase transition in the oxide, known as breakaway effect. In these cracks hydrogen is strongly enriched. It results in very local high hydrogen partial pressure at the oxide/metal interface and in an increase of the hydrogen concentration in the metal at local regions where such cracks in the oxide layer exist. Due to this effect the hydrogen uptake of the remaining zirconium alloy does not depend monotonically on temperature. Differences between (Zr,Sn) and Zr-Nb alloys are caused by differences in the hydrogen production due to different oxidation kinetics and in the crack forming phase transformation in the oxides as well as in the mechanical stability of the oxides.
High-temperature oxidation and quench behaviour of Zircaloy-4 and E110 cladding alloys
2010
This paper gives an overview on the status of knowledge of high-temperature oxidation of the two zirconium alloys Zircaloy-4 and E110 with special emphasis on results obtained during the SARNET period. The tin-bearing alloy Zircaloy-4 and the niobium-bearing alloy E110 are the materials for cladding and structures recently mainly used in pressurised water reactors of the Western type and VVERs and RBMKs, respectively.
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.
Steam and Air Oxidation Behavior of Nuclear Fuel Claddings at Severe Accident Conditions
MRS Proceedings, 2010
The oxidation behavior of zirconium alloys used as materials for nuclear fuel rod claddings is investigated in the temperature range between 973 and 1673 K in steam and air atmosphere. Parabolic kinetics was found for all materials, atmospheres and temperatures, at least at beginning of the reactions. The temperature dependence of the reaction rate is of Arrhenius type. The parameters of the Arrhenius functions are determined and given for steam oxidation. Due to the formation of a large amount of cracks an acceleration of the reactions can occur. Reasons of the crack formations are phase transformations in the oxide layer known as the breakaway effect and, in case of air atmosphere, local oxygen starvation conditions resulting in reactions with nitrogen. The paper gives a short overview of the relevant mechanisms and processes.
Oxygen segregation in pre-hydrided Zircaloy-4 cladding during a simulated LOCA transient
EPJ Nuclear Sciences & Technologies
Oxygen and hydrogen distributions are key elements influencing the residual ductility of zirconiumbased nuclear fuel cladding during the quench phase following a Loss Of Coolant Accident (LOCA). During the high temperature oxidation, a complex partitioning of the alloying elements is observed. A finite-difference code for solving the oxygen diffusion equations has been developed by Institut de Radioprotection et de Sûreté Nucléaire to predict the oxygen profile within the samples. The comparison between the calculations and the experimental results in the mixed a+b region shows that the oxygen diffusion is not accurately predicted by the existing modeling. This work aims at determining the key parameters controlling the average oxygen profile within the sample in the two-phase regions at 1200°C. High temperature steam oxidation tests interrupted by water quench were performed using pre-hydrided Zircaloy-4 samples. Experimental oxygen distribution was measured by Electron Probe Micro-Analysis (EPMA). The phase distributions within the cladding thickness, was measured using image analysis to determine the radial profile of a(O) phase fraction. It is further demonstrated and experimentally checked that the a-phase fraction in these regions follows a diffusion-like radial profile. A new phase fraction modeling is then proposed in the cladding metallic part during steam oxidation. The modeling results are compared to a large set of experiments including the influence of exposure duration and hydrogen content. Another key outcome from this modeling is that oxygen average profile is straightforward derived from the proposed modeling.