Robert J. Comstock

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

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.

Microstructure and Growth Mechanism of Oxide Layers Formed on Zr Alloys Studied with Micro-Beam Synchrotron Radiation

The structures of oxides formed in water and lithiated water on three Zr-based alloys with varied corrosion behavior were studied with micro-beam synchrotron radiation and optical microscopy. Micro-beam synchrotron radiation (0.2 µm spot) has a unique combination of high elemental sensitivity (ppm level) and fine spatial resolution that allowed the determination of various oxide characteristics such as phase content, texture, grain size, and composition as a function of distance from the oxide-metal interface. Micro-beam X-ray fluorescence shows that the oxides formed in lithiated water have increased levels of Fe absorbed from the autoclave environment indicating greater oxide porosity in these oxides. The phase content, texture, and grain size of oxides were studied in detail using synchrotron radiation micro-beam diffraction for samples corroded in water and lithiated water. A remarkable periodicity was observed in the oxide structures using various techniques including X-ray peak intensities for both monoclinic and tetragonal zirconia, texture, and optical microscopy. The periods were similar to the transition period and were less visible in the oxides that behaved worse in lithiated water. These results are discussed in terms of models of oxide growth and of the differences between alloys.

Autoclave study of zirconium alloys with and without hydride rim

Abstract Autoclave corrosion experiments were conducted on a number of zirconium alloys in different heat treatment conditions. The alloys tested in the present work were Zircaloy-4, ZIRLO® (ZIRLO is a registered trademark of Westinghouse Electric Company LLC in the USA and may be registered in other countries throughout the world. All rights reserved. Unauthorised use is strictly prohibited.) and two variants of ZIRLO with significantly lower Sn levels, referred to here as A-0·6Sn and A-0·0Sn. Typical corrosion kinetics with a change from pre- to post-initial transition was observed with ZIRLO and Zircaloy-4 displaying the shortest time to the initial transition after 120–140 days of autoclave exposure, followed by A-0·6Sn materials after 140–260 days. A-0·0Sn materials showed no sign of transition even after 360 days although one sample tested to 540 days had gone through transition. Material in the stress relieved condition generally experienced initial transition earlier than the same alloy in the recrystallised condition. Pretransition samples had a universally black oxide layer, which eventually developed grey patches when transition occurred. Practically, all non-hydrogen charged alloys showed a strong trend towards cubic oxide growth rates. Cathodic hydrogen charging was conducted to simulate end of life condition of cladding tubes, forming a hydride rich rim region at the outer surface of the cladding tubes. Hydrogen charged materials generally experienced accelerated corrosion of different degrees with the exception of recrystallised A-0·0Sn and partially recrystallised A-0·6Sn showing no sign of acceleration. It therefore seems that increasing tin levels has a negative impact on autoclave corrosion behaviour for materials with and without a hydride rich rim. In developing advanced alloys for use in cladding, this effect has been balanced against the benefits that Sn is known to provide in-reactor, including robustness in corrosion behaviour and reduced irradiation growth. It was noted that most materials with a hydride rich rim exhibit parabolic corrosion kinetics with decreased initial weight gain but increased overall weight gain.

ZIRLO TM Cladding Improvement

Improvements to fuel design have been made in recent years to meet the challenges of increases in fuel duty in terms of linear power and operating temperature. Improved cladding material is one of these design improvements. Specifically, Westinghouse has developed an improved version of ZIRLOTM called Optimized ZIRLO and denoted as OPT ZIRLO. The Sn level in this improved material is reduced from the nominal standard previous level of about 1 % to a range of 0.6 % to 0.8 %. The reduced Sn level has been optimized to produce a higher corrosion resistance and provide adequate in-reactor creep resistance. Out-reactor diameter creep tests have shown that decreasing Sn increases out-reactor creep, suggesting that decreasing Sn may also increase in-reactor creep. An in-reactor testing program in the Vogtle Unit 2 pressurized water reactor (PWR) was performed to confirm the predicted in-reactor behavior. The test samples were suspended as segmented rods inside fuel assembly thimble tubes. In-reactor diameter creep data confirmed that decreasing Sn increases in-reactor creep. As a result of the correlation between in-reactor and out-reactor creep, an extensive out-reactor diameter creep program was performed in order to develop methods to fabricate OPT ZIRLO with the same in-reactor creep properties as the currently used stress-relief annealed standard ZIRLO (denoted as SRA STD ZIRLO). The level of in-reactor diameter creep of SRA STD ZIRLO was achieved for OPT ZIRLO by two methods. One method involved changing the final microstructure from SRA to partially recrystallized (PRXA). The other method kept the final microstructure as SRA and changed the tube reduction sequence to decrease the final tube area reduction. In order to develop these methods, a series of material variation tests was performed. Some of the material variations included different final heat treatments, different tube reduction sequences, and different pre-charged hydrogen levels. These tests were performed for both out-reactor and in-reactor. In addition, the out-reactor and in-reactor creep were observed to directly correlate for OPT ZIRLO material fabricated with different final anneal temperatures. Thus, out-reactor creep may be used to predict in-reactor creep properties for different final anneal temperatures. These results show that fabrication changes may be used to control in-reactor creep. In this study, fabrication changes were used to compensate for the reduction in in-reactor creep strength associated with lower tin content in OPT ZIRLO.

Effect of Alloying Elements on Hydrogen Pickup in Zirconium Alloys

Although the optimization of zirconium-based alloys has led to significant improvements in hydrogen pickup and corrosion resistance, the mechanisms by which such alloy improvements occur are still not well understood. In an effort to understand such mechanisms, we conducted a systematic study of the alloy effect on hydrogen pickup, using advanced characterization techniques to rationalize precise measurements of hydrogen pickup. The hydrogen pickup fraction was accurately measured for a specially designed set of commercial and model alloys to investigate the effects of alloying elements, microstructure, and corrosion kinetics on hydrogen uptake. Two different techniques for measuring hydrogen concentrations were used: a destructive technique, vacuum hot extraction, and a non-destructive one, cold neutron prompt gamma activation analysis. The results indicate that hydrogen pickup varies not only from alloy to alloy, but also during the corrosion process for a given alloy. These variations result from the process of charge balance during the corrosion reaction, such that the pickup of hydrogen decreases when the rate of electron transport or Manuscript received December 25, 2012; accepted for publication June 26, 2013; published online June 17, 2014. Dept. of Mechanical and Nuclear Engineering, Penn State Univ., University Park, PA 16802, United States of America (Corresponding author), e-mail: adrien.couet@gmail.com Dept. of Mechanical and Nuclear Engineering, Penn State Univ., University Park, PA 16802, United States of America. Westinghouse Electric Company LLC, Pittsburgh, PA 15235, United States of America. ASTM 17th International Symposium on Zirconium in the Nuclear Industry on February 3–7, 2013 in

Effect of Sn on Corrosion Mechanisms in Advanced Zr-Cladding for Pressurised Water Reactors

The desire to improve the corrosion resistance of Zr cladding material to allow high burnup has resulted in a general trend among fuel manufacturers to develop alloys with reduced levels of Sn. While the detrimental effect of Sn on high temperature aqueous corrosion performance is widely accepted, the reason for it remains unclear. High-Energy synchrotron X-ray diffraction was used to characterise the oxides formed by autoclave exposure on Zr-Sn-Nb alloys with tin concentrations ranging from 0.01 to 0.92 wt.%. The alloys studied included the commercial alloy ZIRLO® and two variants of ZIRLO with significantly lower tin levels, referred to here as A-0.6Sn and A-0.0Sn. The nature of the oxide grown on tube samples from each alloy during autoclave testing at 360°C was investigated by cross-sectional Scanning and Transmission Electron Microscopy (SEM & TEM). Non-destructive synchrotron X-ray diffraction analysis on the oxides revealed that the monoclinic and tetragonal oxide phases display highly compressive in-plane residual stresses with the magnitudes dependent on both phase and alloy. Additional in-situ Synchrotron X-ray diffraction experiments during oxidation at 550°C provided further confirmation of the trends seen for autoclave tested samples and demonstrated the presence of elevated levels of tetragonal phase in the initial stages of oxidation. In-situ and ex-situ measurements demonstrate unambiguously that the amount of tetragonal phase present and, more importantly, the degree of transformation from tetragonal to monoclinic oxide both decrease with decreasing tin levels, suggesting that tin stabilises the tetragonal phase. It is proposed that in Zr-Nb-Sn alloys with low Sn, the tetragonal phase is mainly stabilised by very small grain size and therefore remains stable throughout the corrosion process. By contrast, in alloys with higher tin levels larger, stress stabilised, tetragonal grains can form initially, but then become unstable as the corrosion front progresses inwards and stresses in the existing oxide relax.

Evolution of Texture in Zirconium Alloy Tubing during Processing

Electron backscattered diffraction (EBSD) has been used to determine microtexture in a zirconium alloy tube subject to interrupted pilgering and therefore exhibiting varying amounts of deformation as a function of axial position along the transition between the initial and final size. Texture and hardness measurements have been made as a function of the distance through the tube wall thickness and along the tube length. Texture results have been compared with co-located hardness measurements. The results show a systematic variation in the deformation texture with changes in Q (the ratio of the reduction in thickness to reduction in diameter). This is consistent with previous observations of the effect of Q on texture evolution in zirconium alloys. It is demonstrated that the texture measurements can be correlated well with the anisotropy in strength determined from hardness measurements.

Determination and interpretation of texture evolution during deformation of a zirconium alloy

Worldwide, crystal plasticity models are currently developed to predict texture development during processing of material. Such models require a precise knowledge of the active deformation mechanisms. The activation energy for certain deformation modes will change with temperature and also depend on the chemistry of the alloy as well as the microstructure. Deformation mechanisms were studied in ZIRLO™ during room and high-temperature uniaxial compression testing. Materials with a strong crystallographic basal texture and a more random texture due to β-quenching were investigated with the aim of establishing the effect of temperature, microstructure, and texture on the active deformation modes during the initial stages of deformation. First, specimens were strained at room temperature, 180°C and 300°C to 2 % and 5 % or 10 % total strain and subsequently analyzed by Electron Back Scatter Diffraction (EBSD) to determine the texture evolution. It was found that a dramatic texture change was observed for all testing temperatures in the strongly textured specimen after only 5 % total strain, which can only be understood in terms of tensile twinning of {10 1 ¯ 2 } ⟨ 1 ¯ 0 1 1 ⟩ being active mainly at room temperature and compressive twinning of {11 2 ¯ 2 } ⟨ 1 ¯ 1 ¯ 2 3 ⟩ being operational at room and elevated temperature. The β-quenched specimens did not show any evidence of texture change when strained to 10 %. In-situ intergranular strains were measured by time-of-flight neutron diffraction during continuous compressive loading. This information enabled the development of a crystal plasticity finite element model (CPFEM), which was subsequently used to predict the stress state in individual grains. It was found that in the strongly textured material the spread of intergranular strain in the {0002} grain family (normal pointing towards the ND direction) results in some grains being in compression even though the mean stresses are tensile, which could explain the activation of the observed compressive twinning. The crystal plasticity model also demonstrated that the observed texture changes in the strongly textured material, including those at high temperature, cannot be explained by slip alone even when ⟨c+a⟩ slip is considered. In addition, the model showed that the dramatic difference in yield strength of the two conditions studied here cannot be solely attributed to the difference in texture but that grain size plays an important role.

Texture Evolution of Zircaloy-2 During Beta-Quenching: Effect of Process Variables

The nuclear industry is interested in developing thermomechanical processes to produce random crystallographic orientation (texture) from cold-rolled Zircaloy-2 sheets used to manufacture boiling water reactor (BWR) channels. Randomized textures are beneficial because they minimize anisotropic irradiation-assisted growth, which in turn reduces bowing and uncontrolled deformation of BWR channels during service. The texture evolution of cold-rolled Zircaloy-2 sheets during the allotropic α→β→α phase transformation was characterized by using synchrotron X-ray diffraction in situ and electron backscatter diffraction. The initial strong rolling texture is weakened only if the α→β→α phase transformation is complete. Plastic deformation and grain growth in the β-phase lead to changes in the β texture and modify the inherited α texture. The global texture evolution is not sensitive to levels of stress that do not cause β plastic deformation. These findings demonstrate that accurate temperature control of the β-quenching process is of utmost importance in order to minimize undesirable irradiation growth of BWR channels during service, and that plastic deformation in the β phase can be employed to modify the inherited α texture. BWR channels with β-quenched textures will exhibit minimum irradiation growth caused by texture.

Application of modeling of the texture dependence of environmentally assisted crack growth of long and short cracks to ZIRCALOY fuel tubing

Argon and iodine stress-rupture tests were performed on five lots of ZIRCALOY-4 tubing with relatively large differences in texture. The addition of iodine, relative to argon, decreased the failure time. The iodine data exhibited increasing failure times with decreasing stress until a plateau or threshold stress level was reached. The threshold stress was used to evaluate a model developed from fracture mechanics crack propagation data. Modification of this model was necessary in order to account for tubing texture, tubing fracture surface characteristics, test temperature, and the embrittling effect of iodine. The adjusted model is consistent with the experimental data and predicts that moderate increases in the iodine threshold stress may be obtained with very low tangetial texture tubing.