Relevant bibliographies by topics / Zirconium alloys – Corrosion

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  1. Journal articles
  2. Dissertations / Theses
  3. Books
  4. Book chapters
  5. Conference papers
  6. Reports

Academic literature on the topic 'Zirconium alloys – Corrosion'

Author: Grafiati
Published: 4 June 2021
Last updated: 10 February 2022

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Journal articles on the topic "Zirconium alloys – Corrosion"

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Motta, Arthur T. "Waterside corrosion in zirconium alloys." JOM 63, no. 8 (August 2011): 59–63. http://dx.doi.org/10.1007/s11837-011-0140-0.

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Cox, B. "Stress corrosion cracking of zirconium alloys." Langmuir 3, no. 6 (November 1987): 867–73. http://dx.doi.org/10.1021/la00078a002.

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Kim, Minsuk, Seongbin An, Chaeeul Huh, and Chungseok Kim. "Development of Zirconium-Based Alloys with Low Elastic Modulus for Dental Implant Materials." Applied Sciences 9, no. 24 (December 4, 2019): 5281. http://dx.doi.org/10.3390/app9245281.

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The stress-shielding effect is a phenomenon in which the mutual coupling between bones and bio-materials of the human body is loosened due to the difference in elastic modulus, and bone absorption occurs due to the difference in density, which causes a shortening of the life of the material. The purpose of this study is to develop a zirconium-based alloy with low modulus and to prevent the stress-shielding effect. Zr–7Cu–xSn (x = 1, 5, 10, 15 mass%) alloys were prepared by an arc-melting process of pure zirconium, oxygen-free copper, and tin, respectively. The Zr–7Cu–xSn alloy has two phase α-Zr and Zr2Cu intermetallic compounds. Microstructure characterization was analyzed by microscopy and X-ray diffraction. Corrosion tests of zirconium-based alloys were conducted through polarization tests, and zirconium-based alloys had better corrosion characteristics than other metal bio-materials. In general, the elastic modulus value (14–25 GPa) of the zirconium-based alloy is very similar to the elastic modulus value (15–30 GPa) of the human bone. Consequently, the zirconium-based alloy is likely to be used as a bio-material that negates the effect of stress shielding on human bones.
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dos Santos, Demetrio Jackson, Lara Basilio Tavares, and Maria Cecilia Salvadori. "Zirconium Based Metal Pretreatments: A Characterization Method for Ecologically Sustainable Thin Film Surface Pretreatments." Materials Science Forum 869 (August 2016): 693–98. http://dx.doi.org/10.4028/www.scientific.net/msf.869.693.

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Zirconium based metal pretreatments have become widely used in recent years as a substitute for phosphate deposition on steel alloys and for chromate on aluminum alloys in industrial applications. The choice of a zirconium based intermediate layer follows from its ecologic sustainability − decreased water and energy consumption, vehicle weight reduction, and low byproduct generation during processing. Here we describe our investigations of a characterization method of converted metal oxide thin films deposited by a plasma method. The thin film composition was characterized by Rutherford Backscattering Spectroscopy (RBS) and Energy Dispersive Spectroscopy (EDS) before and after conversion by a zirconium-based pretreatment, revealing the formation of zirconia after treatment. The corrosion mechanism of the deposited metal oxide films was investigated using electrochemical analysis, confirming the susceptibility of the film to corrosion and the applicability of corrosion investigations. The results pointed to a better performance of the RBS in comparison to EDS.
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Mousavian, Seyed Mohammad Hossein, Seyed Hadi Tabaian, and Mohammadhassan Badihehaghdam. "Effects of zirconium addition on electrochemical and mechanical properties of Mg-3Zn-1Ca-1RE alloy." Anti-Corrosion Methods and Materials 67, no. 6 (November 26, 2020): 583–91. http://dx.doi.org/10.1108/acmm-06-2020-2324.

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Purpose The effect of zirconium, zinc, calcium and rare earth group as the alloying elements on mechanical properties and corrosion behavior of magnesium alloys was investigated in the simulated body fluid. Design/methodology/approach Pure magnesium and the alloying elements were melted and zirconium was finally added to obtain different alloys. The castings were annealed and some samples were aged heat treated. X-ray fluorescence was used for the elemental analysis and LSV was used for electrochemical corrosion evaluations. Findings Results showed that corrosion resistance decreases with increasing zirconium content. The lowest corrosion rate was obtained for the samples containing 0.3% and 0.45% of Zr from annealed and aging heat-treated samples, respectively. Yield stress enhances with increasing the zirconium content and degrades by the aging heat treatment. Originality/value These alloys were studied for the first time. Effect of casting without using protective flux and vacuum furnaces. Effect of annealing at 440°C for 2 h and artificial aging at 200°C for 16 h. Alloy’s electrochemical behavior on the body’s simulation environment has been investigated. Improvement of mechanical properties after annealing heat treatment by high zirconium percentage.
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Zhang, Jun Song, Chong Sheng Long, Jing Jing Liao, Tian Guo Wei, and Zhong Bo Yang. "Deoxidation Process of Oxidized Zirconium Alloy." Materials Science Forum 993 (May 2020): 22–28. http://dx.doi.org/10.4028/www.scientific.net/msf.993.22.

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When zirconium alloy is corroded, an oxide film is formed on the surface, which hinders the ion transfer during the corrosion process. Therefore, the analysis of the oxide film is an important part of the research on the corrosion resistance of zirconium alloys. In this paper, two kinds of Zr-Sn-Nb alloys were corroded in 400 °C/10.3 MPa pure steam and 500 °C/10.3 MPa pure steam in autoclave to obtain samples with oxide thickness of 14 um and 18 um respectively. Then they were annealed at 800 °C at a pressure of 10-4 Pa for 18 h. XRD and WDS studies were used to analyze the structure and oxygen content of the oxide film after annealing. The results indicate that the oxide films of alloys change from zirconium dioxide to zirconium after annealing. The oxygen diffuses into the substrate and its content decreases continuously with increasing diffusion distance. Combined with the SEM analysis of cross-section samples, it is found that the annealed samples are composed of several layers. An oxygen-saturated zirconium layer, a transitional layer with micro-cracks, an oxygen-dissolved α-Zr layer and a β-Zr layer are identified. Based on these results, the mechanism of the ion transfer in the oxide film during annealing is analyzed deeply. It is proposed that space charges in the oxide film have a major impact on deoxidation kinetics. This study provides a new research method for the corrosion mechanism of zirconium alloys.
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Jithendra Kumar, Repalle, B. V. S. Raghu Vamsi, T. Siva Krishna, D. Tarun, and M. Kamal Tej. "Evaluation of Coefficient of Thermal Expansion of Zirconium by Using Dilatometer & Ansys." Advanced Materials Research 1148 (June 2018): 128–35. http://dx.doi.org/10.4028/www.scientific.net/amr.1148.128.

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Zirconium alloys are solid solutions of zirconium or other metals. Zirconium has very low absorption cross-section of thermal neutrons. Zirconium has high corrosion resistance, ductility and hardness. Zirconium is mainly used as a good refractory metal. Zirconium can be manufactured by using standard fabrication techniques. In the present scenario zirconium alloys are used in water reactors for the cladding of fuel rods in nuclear reactors in nuclear technology. We use the composition of zirconium alloys as more than 94.5 weight percentage of zirconium and less than 2.45 weight percentage of copper which are added to improve mechanical, thermal properties and corrosion resistance. This paper first focuses on the study of thermal properties of Zirconium. And this particularly concentrated on variation of Coefficient of Thermal Expansion by varying temperatures by using Dilatometer and as well as ANSYS
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Preuss, M., P. Frankel, S. Lozano-Perez, D. Hudson, E. Polatidis, N. Ni, J. Wei, et al. "Studies Regarding Corrosion Mechanisms in Zirconium Alloys." Journal of ASTM International 8, no. 9 (2011): 103246. http://dx.doi.org/10.1520/jai103246.

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Tsutsumi, Yusuke, Yousuke Takano, Hisashi Doi, Kazuhiko Noda, and Takao Hanawa. "Corrosion Behavior of Zirconium Based Alloys in Simulated Body Fluids." Materials Science Forum 561-565 (October 2007): 1489–92. http://dx.doi.org/10.4028/www.scientific.net/msf.561-565.1489.

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The corrosion behavior of Zr-Ti and Zr-Hf alloys with various concentrations in properly deaerated Hanks’ solution were investigated by electrochemical techniques. The effects of the thermal treatment with the purpose of homogenization of chemical composition on the pitting potentials of the Zr-Ti alloys were also examined. The results indicated that sufficient addition of Ti to Zr improved its corrosion resistance, especially in terms of the pitting potential and the passive current density. On the other hand, addition of Hf totally lowered the corrosion resistance of the alloy. The thermal treatment improved the pitting corrosion property of Zr-Ti alloys, and Zr alloys with over 5mol% Ti showed much higher pitting potentials than that of pure Zr. The treatment also improved the reproducibility of the measurement and narrowed the data scattering. This phenomenon was discussed with the model of rapidly-cooled metallographic structure.
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Luo, J. S., and D. P. Abraham. "Transmission Electron Microscopy of Corrosion of Stainless Steel-Zirconium Metal Waste Forms." Microscopy and Microanalysis 5, S2 (August 1999): 848–49. http://dx.doi.org/10.1017/s1431927600017566.

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Stainless steel-zirconium (SS-Zr) alloys have been developed as waste forms to immobilize and retain fission products generated during the electrometallurgical treatment of spent nuclear fuel. The baseline waste form is a stainless steel-15 wt.% zirconium (SS-15Zr) alloy, which is prepared by melting appropriate amount of Type 316 stainless steel (SS316) and high purity zirconium. As zirconium has very low solubility in iron, the addition of zirconium to SS316 results in the formation of ZrFe2 -type Laves intermetallic phases. The corrosion behavior of stainless steel has been widely studied; however, the corrosion behavior of the Zr-based-intermetallic has not been previously investigated. In this paper, we present a microstructural characterization of the corrosion layer formed on the Zr-intermetallic phase using energy-filtering transmission electron microscopy (TEM) and energy dispersive x-ray spectroscopy (EDS).Specimens of SS-15Zr alloy, crushed to 75 to 150 μm sizes, were immersed in 90°C deionized water for a period of two years.
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Dissertations / Theses on the topic "Zirconium alloys – Corrosion"

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Treeman, Nicole M. "Electrochemical study of corrosion phenomena in zirconium alloys." Thesis, Cambridge, Massachusetts, Massachusetts Institute of Technology, 2005. http://hdl.handle.net/10945/37785.

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CIVINS
Shadow corrosion of zirconium alloy fuel cladding in BWR environments, the phenomenon in which accelerated corrosion is experienced when the cladding surface is in close proximity to other metals, has become a potential life-limiting issue for BWR fuel. Recent results from experimentation at MIT, Halden and Studvik suggest that a galvanic coupling drives the phenomenon between the cladding and the adjacent material. However, the actual processes involved are not understood. One key parameter that would help in the understanding of the phenonenon would be a measurement of the actual corrosion current between fuel cladding and adjacent materials in the actual in-reactor environment. The limitations placed on the burn-up of uranium oxide fuel corelates to the amount of corrosion seen through a directly measurable oxide thickness on the waterside of the zirconium alloy cladding. This oxide corrosion product directly correlates to distance from structure components, leading the effect commonly referred to as shadow corrosion. In recent experiments, Studvik determined that there are large ECP differences associated with Inconel and zirconium alloys that correlate to increased galvanic current density when the materials are coupled. In this thesis research, four electrode pairs wre used to measure galvanic cirrent densities in the irradiation environment:Pt-Pt, Zircalloy (Zr-2), Inconel (X-750)-Pt, and Zr-2-X-750. To determine the changes in the coolant water conductivity dus to the presence of radiolysis products, electrochemical potential mesurements of Pt-Pt coupled electrodes were analyzed. Finally, attempts to characterize the observed oxide behavior using measurement from Electrochemical Impedance Spectroscopy (EIS), also known as Alternating Current Impedance, were conducted. Through the measurements taken, analysis of the mechanisms potentially causing the shadow corrosion phenomenon was conducted. The results of the observations included: Measurement of increased conductivity of coolant water correlating to increases in reactor power, measurement of increased galvanic current measurements correlating to increases in reactor power.
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Setiadinata, Sylvester Brian. "Corrosion and hydrogen pickup mechanisms of zirconium alloys." Thesis, University of Oxford, 2016. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.728802.

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Zhang, Yue. "Corrosion of titanium, zirconium and their alloys for biomedical applications." Thesis, University of Birmingham, 2018. http://etheses.bham.ac.uk//id/eprint/8033/.

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in vivo degradation of Ti implants has remained as a concern despite its perceived excellent corrosion resistance. Elevated levels of metal have been detected both locally and systemically, often leading to unfavourable biological responses. Corrosion of Ti, Zr and their alloys (Ti6Al4V and TiZr) were studied in simulated physiological solutions. It was found that albumin, the most prevalent tissue fluid protein, induced a time dependent dissolution of Ti6Al4V in the presence of H2O2, an inflammatory biomolecule commonly found in peri implant sites, in 0.9% NaCl. However, the corrosion of Zr was observed to be unaffected by the presence of H2O2 and/or albumin in 0.9% NaCl. Furthermore, TiZr alloys have been shown to possess enhanced passivity in comparison to CP Ti in the various exposure conditions including highly acidic (HCl), oxidative environments (H2O2 in 0.9% NaCl) and cell cultures (macrophage). Corrosion products of Zr were characterised in situ by synchrotron X ray methods, which were found to be ZrOCl2 ∙ 8H2O, tetragonal ZrO2, and Zr metal fragments in 0.9% NaCl regardless of the presence or absence of H2O2 and/or albumin. The presence of Zr metal fragments as a result of a corrosion process indicates the generation of metal species in the absence of wear.
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Wei, Jianfei. "Effect of hydrogen on the corrosion performance of zirconium alloys." Thesis, University of Manchester, 2012. https://www.research.manchester.ac.uk/portal/en/theses/abstracteffect-of-hydrogen-on-the-corrosion-performance-of-zirconium-alloys(b6683f8b-9286-4ebd-bf8a-d240c62b7a79).html.

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This study is part of the multidisciplinary research program funded by the EPSRC aimed at further understanding the mechanisms behind the oxidation, transition and breakaway of zirconium alloys. The current study concentrates on the effect of artificially introduced hydride rich rim region upon the corrosion performance of zirconium alloys. The methodology of cathodic hydrogen charging of zirconium specimen has been described. The intention is to create a hydride rich rim similar to observation made in end-of-life fuel claddings from nuclear reactor. The overall hydrogen levels in the samples were determined via inert gas fusion and the local hydride volume fractions via laboratory x-ray diffraction. The residual stress state in the rim was investigated by synchrotron x-ray diffraction. The rim was determined to have local hydrogen levels about 11,000 wt.ppm and predominantly δ-hydrides. Hydrided samples and their references were tested in autoclaves simulating primary water environment. Reference samples exhibited typical corrosion kinetics with a change from pre- to post-1st transition. The duration before 1st transition varies with alloy chemistry and heat treatment conditions. Hydrided samples generally experienced accelerated corrosion while the extent to which the material was affected depends on alloy chemistry. Parabolic rather than cubic oxide growth with a very slow early corrosion rate was recorded in hydrided alloys. Both hydrided and reference samples of three Zr-Sn-Nb alloys, ZIRLOTM, low tin ZIRLO and X2 with tin concentration ranging from 0.14 to 0.92 wt.% became the gravity of further characterisations since they exhibited improved corrosion performance with decreasing tin content. The microstructure of the oxide was investigated via cross-sectional Scanning Electron Microscope (SEM). The residual stress state in tetragonal and monoclinic phases as well as tetragonal phase fraction were studied via synchrotron x-ray diffraction, showing high in-plane compressive stresses in the oxide of different levels depending on the phase and alloy. The presence of a hydride rim lowered the residual stress. White oxide in hydrided samples displays lower residual stress level, smaller tetragonal phase fraction and higher degree of oxide grain misorientation when compared to black oxide regions of reference and hydrided samples. It was also found that the hydride rim region undergoes local depletion of hydrides in alloy with low Sn content during the early stage of aqueous corrosion. Finally a model was given to describe the different roles played by individual oxide and hydride phases during the oxidation of zirconium alloys.
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Feltham, Andrew Martin Charles. "The corrosion and passivity of some engineering alloys in acidic solution." Thesis, University of Cambridge, 1990. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.359526.

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Garner, Alistair John. "Investigating the effect of oxide texture on the corrosion performance of zirconium alloys." Thesis, University of Manchester, 2015. https://www.research.manchester.ac.uk/portal/en/theses/investigating-the-effect-of-oxide-texture-on-the-corrosion-performance-of-zirconium-alloys(c98f3395-4ee7-42c5-b9f3-2a55e7923c14).html.

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This work was performed as part of the MUZIC-2 (Mechanical Understanding of Zirconium Corrosion) collaboration, established with the goal of understanding the mechanism of hydrogen pickup in zirconium alloys. Hydrogen pickup is one of the least understood and most significant degradation mechanisms affecting zirconium alloys in nuclear reactors. These alloys are used as cladding and structural materials in the reactor core, mainly due to their low thermal neutron absorption cross section and excellent corrosion resistance. This project aims to investigate the effect of oxide texture (i.e. the degree of preferred orientation) on the corrosion performance of zirconium alloys. The texture of the oxide is expected to affect the microstructural development of the oxide, the grain boundary distribution and the stress state. It is therefore considered to be one of the most important factors in determining how the corrosion process occurs, and why different alloys exhibit significantly different corrosion performance. It is hoped that this project will add to the current knowledge of the corrosion process, and in particular hydrogen pickup, so that the route of hydrogen through the protective oxide can be identified. This will lead to the development of a new generation of alloys that provide improved oxidation and hydrogen pickup performance, whilst maintaining the required mechanical properties. This work focuses on four zirconium alloys; Zircaloy-4, ZIRLO™, low-Sn ZIRLO™ and Zr-1.0Nb-0.1Fe. The alloys all have different chemical compositions and therefore exhibit different corrosion performance. The macrotexture of the oxide formed on different alloys was measured by glancing angle X-ray diffraction (XRD) and Electron Backscatter Diffraction (EBSD). A fibre texture was formed in all cases, with the (10-3) to (10-5) planes oriented parallel to the metal-oxide interface for the monoclinic phase. The major orientation was found to be independent of alloy chemistry, substrate orientation and oxidation conditions. The monoclinic texture strength was found to be weakened with increasing oxidation temperature, The major orientation of the tetragonal phase was also found to be a fibre texture, with the (001) planes oriented approximately parallel with the interface. Although significant variation from this texture component was observed. It is suggested that the main driving force for oxide texture development is the transformation stress induced by the Zr-ZrO2 transformation. The microtexture of the oxides was measured using two novel techniques, Transmission Kikuchi Diffraction (TKD) and automated crystal orientation mapping with transmission electron microscopy (TEM). The techniques revealed an oxide microstructure consisting of an outer layer of equiaxed grains with a large range of orientations and an inner layer consisting of well-aligned columnar monoclinic grains. This layer of equiaxed grains was observed to form again after the transition in corrosion kinetics. A large fraction of transformation twin boundaries were observed throughout the oxides. Some of these boundaries were observed to surround large monoclinic grains, providing evidence that the tetragonal grains from which they transformed had grown considerably larger than the critical grain size for stabilisation. Without the presence of Sn, larger monoclinic grains were observed to form with a greater degree of preferred orientation, and with a lower fraction of transformation twin boundaries than Sn-containing alloys. In addition, an increased number of well-oriented tetragonal grains was correlated with the presence of Sn. It is therefore concluded that a reduction in Sn will lead to the formation of an oxide microstructure that is more resistant to both oxidation and hydrogen pickup. An analysis of the crystallography and morphology of a ZrO phase present at the metal-oxide interface was also performed. Finally, the observations have been used to construct a model of the corrosion process.
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Châtelain, Anthony R. (Anthony Roger) 1972. "Enhanced corrosion of zirconium-based alloys in proximity to other metals : the "shadow effect"." Thesis, Massachusetts Institute of Technology, 2000. http://hdl.handle.net/1721.1/8871.

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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2000.
Includes bibliographical references.
Fuel cladding for water-cooled power reactors must meet certain requirements for optimal performance. To function in the extreme conditions typical of a nuclear reactor core the material used must be corrosion resistant, have low thermal neutron cross section, and high strength. Corrosion resistance is one of the most important parameters for reactor materials. From the beginning of the use of reactors, engineers have been faced with the problem of excessive corrosion in several different forms. In recent years, a peculiar corrosion phenomenon has increased in significance. Several occurrences of local corrosion enhancement of zirconium-base alloys in proximity to other components have been observed. This corrosion enhancement talcs the form of a "shadow" of a metal component in proximity, hence its name, "shadow effect." Although much recent attention has been given to the shadow effect, it has been known since the sixties, but has only lately been considered a possible threat to material integrity. Today the interest in local corrosion enhancement due to the shadow effect and its implications for in-core performance of cladding and structural material is increasing worldwide. International experience has shown that the phenomenon has occasionally resulted in serious corrosion problems threatening material integrity. In order to prevent future obstruction from the phenomenon an understanding of the shadow effect needs to be developed. This becomes important in today's rapid expansion of aggressive reactor environments with higher burn-up and the need for longer fuel residence times for more economical runs. This project was conducted at MIT, funded by ABB Atom, which had the goal of identifying the basic mechanisms of the shadow effect. The MIT research reactor MITR-11 was used to simulate BWR core coolant conditions. The sample train included Zr-2- alloy with various surface treatments. Different counter electrodes surrounded each cladding piece. They were high and low beta emitters, inert material and Zircaloy-2 in contact and non-contact at various separation distances. Post-irradiation examination of the cladding pieces showed: * Beta-radiation is not the main mechanism for the shadow effect. * Shadow corrosion is partly dominated by an electrochemical mechanism. * Radiolysis plays an important role for the formation of shadow corrosion.
by Anthony R. Châtelain.
S.M.
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Wang, Peng. "Corrosion behaviour of zirconium alloys in high temperature aqueous environment by electrochemical impedance spectroscopy." Thesis, University of Manchester, 2011. https://www.research.manchester.ac.uk/portal/en/theses/corrosion-behaviour-of-zirconium-alloys-in-high-temperature-aqueous-environment-by-electrochemical-impedance-spectroscopy(e1bf6a9f-c8ca-45db-8e05-14ee723886d9).html.

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The corrosion behaviour of zirconium based alloys has been primarily investigated by electrochemical impedance spectroscopy (EIS). In-situ autoclave EIS experiments were performed in simulated primary coolant conditions in order to study the high temperature water corrosion of zirconium alloys in PWRs. In-situ impedance response of the corroding material was recorded throughout first kinetic transition. A physical model of the zirconium oxide was proposed in accordance with the microstructural observation' made by SEM analysis. Electrical properties of the oxide was evaluated with equivalent circuit model (ECM) which was constructed according to the physical oxide model. Evolution of various oxide parameters obtained from ECM was analysed in accordance with the microstructure observation made by SEM. A two layer structure consists of a outer porous oxide and an inner barrier oxide, was found to be the most accurate description for the autoclave formed oxide. Supporting evidence from the SEM cross-section and surface analysis of the oxide had shown cracks and pores that were linked and connected with the environment. This observation is also confirmed by the in-situ EIS measurement which has shown porous electrode behaviour throughout the course of oxidation. The porous oxide behaviour was also confirmed by the ex-situ soaking experiment on samples with incremental exposure time. Evolution of inner barrier layer oxide thickness was found to be correlated with kinetic transition which was determined from weight gain measurement. This indicated that barrier layer maybe the oxidation rate controlling layer and its thickness maybe reduced during transition. Thus, a thinner barrier layer would resulted in a rapid corrosion of zirconium alloys. Furthermore, maintaining the barrier layer thickness maybe the possible route to improve zirconium alloy corrosion resistance.
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Lalgudi, Srinivas Bhadrinarayanan. "Experimental evalution of oxide growth in binary zirconium alloys along with pure zirconium and Zircaloy-2 by steam corrosion testing." Thesis, KTH, Fysik, 2012. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-95310.

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Proff, Christian. "Aspects microstructuraux de l'oxydation d'alliages de Zirconium." Phd thesis, Université de Grenoble, 2011. http://tel.archives-ouvertes.fr/tel-00609232.

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Cette thèse est axée sur la caractérisation microstructurale des précipités dans les oxydes des alliages binaires de zirconium (1 wt.% Fe, Cr , Ni ou 0.6 wt.% Nb). La température d'oxydation est fixée au 415°C. Les échantillons sont oxydés dans l'air et dans l'autoclave sous des pressions différentes et dans un microscope électronique à balayages environnemental sous vapeur d'eau. Les résultats des recherches peuvent être résumés ci-dessous : -Deux types d'oxydation (retardée et non retardée) ont été observés pour les précipités. -Le facteur de Pilling-Bedworth des précipités est plus élevé par rapport à celui de zirconium. -Les précipités contenant du fer entrainent une formation des cristaux de l'oxyde de fer pur à la surface du matériau, quand les précipités sont à la surface ou à la proximité. Ces observations mènent à la conclusion que le comportement d'oxydation des précipités peut être corrélé à leurs compositions et à la tendance d'oxydation de leurs éléments constituants.
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Books on the topic "Zirconium alloys – Corrosion"

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Khatamian, D. Hydrogen traps in the oxide/alloy interface region of Zr-Nb alloys. Chalk River, Ont: Reactor Materials Research Branch, Chalk River Laboratories, 1995.

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Schaefer, R. J. Interaction of zinc vapor with Zircaloy and the effect of zinc vapor on the mechanical properties of zircaloy. Washington, DC: U.S. Nuclear Regulatory Commission, 2000.

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Tsai, H. Zinc-zircaloy interaction in dry storage casks. Washington, DC: U.S. Nuclear Regulatory Commission, 2001.

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Uziemblo, Nancy Helen. Use of a pH stat system in the measurement of Zircaloy-4 corrosion in the presence of fluoride. 1989.

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Reitz, W. Influence of laser processing on the corrosion and microstructure of zirconium based material. 1990.

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National Institute of Standards and Technology (U.S.), ed. The reactions of zinc vapor with Zircaloy-4 and pure zirconium. Gaithersburg, MD: U.S. Dept. of Commerce, Technology Administration, National Institute of Standards and Technology, 2000.

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Interaction of zinc vapor with Zircaloy and the effect of zinc vapor on the mechanical properties of Zircaloy. Washington, DC: Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 2000.

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VanWinkle, James A. Oxidation of Zircaloy-4 in pressurized water reactor conditions. 1989.

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VanWinkle, James A. Oxidation of Zircaloy-4 in pressurized water reactor conditions. 1989.

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General Motors Research and Development Center. and United States. National Aeronautics and Space Administration., eds. Spitfire-1: A cooperative agreement for the development of rapid low-cost superplastic forming of aluminum : progress report, twelve-month deliverables, March 1, 1995-May 31, 1995. Warren, MI: The Center, 1995.

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Book chapters on the topic "Zirconium alloys – Corrosion"

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Preuss, M., P. Frankel, S. Lozano-Perez, D. Hudson, E. Polatidis, N. Ni, J. Wei, et al. "Studies Regarding Corrosion Mechanisms in Zirconium Alloys." In Zirconium in the Nuclear Industry: 16th International Symposium, 649–81. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2011. http://dx.doi.org/10.1520/stp49278t.

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Preuss, M., P. Frankel, S. Lozano-Perez, D. Hudson, E. Polatidis, N. Ni, J. Wei, et al. "Studies Regarding Corrosion Mechanisms in Zirconium Alloys." In Zirconium in the Nuclear Industry: 16th International Symposium, 649–81. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2011. http://dx.doi.org/10.1520/stp49376s.

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Preuss, M., P. Frankel, S. Lozano-Perez, D. Hudson, E. Polatidis, N. Ni, J. Wei, et al. "Studies Regarding Corrosion Mechanisms in Zirconium Alloys." In Zirconium in the Nuclear Industry: 16th International Symposium, 649–81. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2011. http://dx.doi.org/10.1520/stp152920120026.

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Markelov, Vladimir, Vladimir Novikov, Alexandr Shevyakov, Anatoly Gusev, Michael Peregud, Viktor Konkov, Sergey Eremin, Alexander Pokrovsky, and Alexander Obukhov. "Preliminary Irradiation Effect on Corrosion Resistance of Zirconium Alloys." In Zirconium in the Nuclear Industry: 18th International Symposium, 857–80. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2018. http://dx.doi.org/10.1520/stp159720160067.

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Cox, B. "Modelling the Corrosion of Zirconium Alloys in Nuclear Reactors Cooled by High Temperature Water." In Modelling Aqueous Corrosion, 183–200. Dordrecht: Springer Netherlands, 1994. http://dx.doi.org/10.1007/978-94-011-1176-8_9.

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Tsutsumi, Yusuke, Yousuke Takano, Hisashi Doi, Kazuhiko Noda, and Takao Hanawa. "Corrosion Behavior of Zirconium Based Alloys in Simulated Body Fluids." In Materials Science Forum, 1489–92. Stafa: Trans Tech Publications Ltd., 2007. http://dx.doi.org/10.4028/0-87849-462-6.1489.

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Tupin, Marc, Joel Hamann, Damien Cuisinier, Philippe Bossis, Martine Blat, Antoine Ambard, Alain Miquet, Damien Kaczorowski, and François Jomard. "Understanding of Corrosion Mechanisms of Zirconium Alloys after Irradiation: Effect of Ion Irradiation of the Oxide Layers on the Corrosion Rate." In Zirconium in the Nuclear Industry: 17th Volume, 438–78. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2014. http://dx.doi.org/10.1520/stp154320120199.

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Yu, Zefeng, Michael Moorehead, Léo Borrel, Jing Hu, Mukesh Bachhav, and Adrien Couet. "Fundamental Understanding of Nb Effect on Corrosion Mechanisms of Irradiated Zr-Nb Alloys." In Zirconium in the Nuclear Industry: 19th International Symposium, 669–95. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2021. http://dx.doi.org/10.1520/stp162220190021.

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Franklin, David G. "Performance of Zirconium Alloys in Light Water Reactors with a Review of Nodular Corrosion." In Zirconium in the Nuclear Industry: 16th International Symposium, 17–36. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2010. http://dx.doi.org/10.1520/stp49254t.

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Franklin, David G. "Performance of Zirconium Alloys in Light Water Reactors with a Review of Nodular Corrosion." In Zirconium in the Nuclear Industry: 16th International Symposium, 17–36. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2010. http://dx.doi.org/10.1520/stp49352s.

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Conference papers on the topic "Zirconium alloys – Corrosion"

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Yao, M. Y., B. X. Zhou, Q. Li, W. P. Zhang, L. Zhu, L. H. Zou, J. L. Zhang, and J. C. Peng. "Effect of Bi Addition on the Corrosion Behavior of Zirconium Alloys." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-15460.

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In order to investigate systematically the effect of Bi addition on the corrosion resistance of zirconium alloys, different zirconium-based alloys, including Zr-4 (Zr-1.5Sn-0.2Fe-0.1Cr), S5 (Zr-0.8Sn-0.35Nb-0.4Fe-0.1Cr), T5 (Zr-0.7Sn-1.0Nb-0.3Fe-0.1Cr) and Zr-1Nb, were adopted to prepare the zirconium alloys containing Bi of 0∼0.5% in mass fraction. These alloys were denoted as Zr-4+xBi, S5+xBi, T5+xBi and Zr-1Nb+xBi, respectively. The corrosion behavior of these specimens was investigated by autoclave testing in lithiated water with 0.01 M LiOH or deionized water at 360°C/18.6 MPa and in superheated steam at 400 °C/10.3 MPa. The micro structure of the alloys was examined by TEM and the second phase particles (SPPs) were analyzed by EDS. Micro structure observation shows that the addition of Bi promotes the precipitation of Sn as second phase particles (SPPs) because Sn is in solid solution in α-Zr matrix in Zr-4, S5 and T5 alloys. The concentration of Bi dissolved in α-Zr matrix increase with the increase of Nb in the alloys, and the excess Bi precipitates as Bi-containing SPPs. The corrosion results show that the effect of Bi addition on the corrosion behavior of different zirconium-based alloys is very complicated, depending on their compositions and corrosion conditions. In the case of higher Bi concentration in α-Zr, the zirconium alloys exhibit better corrosion resistance. However, in the case of precipitation of Bi-containing SPPs, the corrosion resistance gets worse. This indicates that the solid solution of Bi in α-Zr matrix can improve the corrosion resistance, while the precipitation of the Bi-containing SPPs is harmful to the corrosion resistance.
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Zeng, Qifeng, Libing Zhu, Gaihuan Yuan, Lian Wang, Guocheng Sun, and Junqiang Lu. "Microstructure and Properties of New Zirconium Alloys for CAP1400 Fuel Assembly." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-66951.

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Two Zr-Sn-Nb alloys with minor Germanium or silicon additions were prepared by traditional manufacturing process to meet the design requirements. Transmission electron microscope (TEM) and scanning electron microscope (SEM) were utilized to characterize the detail microstructure of base alloys. Corrosion resistance was examined by the weight gain in static autoclave with different water chemistry environments. The mechanical properties at room temperature and elevated temperature were evaluated by conventional tensile testing. Thermal creep resistance was evaluated by an internally-pressurized creep test at 385 °C with hoop stresses of about 108 MPa and 150 MPa (during 24 h). It was found that SZA-6 and SZA-4 alloys consisted of partially recrystallized grain structures with uniformly distributed fine second phase particles (SPPs) located within grain interior and at grain boundaries. Both SZA-4 and SZA-6 alloys exhibited excellent corrosion resistance in two water chemistry conditions. The corrosion resistance of SZA-6 was better than the reference commercial alloy, and SZA-4 was slightly better than SZA-6. The mechanical properties of two new zirconium alloys were comparable, and both of them can meet the design criterion. Moreover, the thermal creep resistance of SZA-4 and SZA-6 alloys was equivalent to existing commercial alloy. Considering the outstanding corrosion resistance, satisfied mechanical properties and thermal creep resistance, SZA-4 and SZA-6 alloys were suggested as promising alloys used for CAP1400 fuel assembly in the future.
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Škarohlíd, Jan, Radek Škoda, and Irena Kratochvílová. "High Temperature Oxidation of Polycrystalline Diamond Coated Zirconium Alloy." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60596.

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Polycrystalline diamond coating is a promising possibility for prevention, or reduction of high temperature oxidation of zirconium alloys and decrease corrosion rate of zirconium alloy during standard operation. Zirconium alloys are widely used as cladding and construction material in almost all types of nuclear reactors, where usually creates a barrier between nuclear fuel and cooling water in the primary circuit. Hydrogen and considerable amount of heat is released during steam oxidation that may occur in an eventual accident. In this paper zirconium alloy was covered by polycrystalline diamond layer using Plasma Enhanced Linear Antennas Microwave Chemical Vapor Deposition system reactor. X-Ray Diffraction and Raman spectroscopy measurements confirmed coverage of the surface area with crystalline and amorphous carbon layer. Characterizations (Raman spectroscopy) were done for zirconium alloy covered with polycrystalline diamond layer before and after high temperature steam exposure. Weight increase and hydrogen release ware measured during steam exposure.
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Škarohlíd, Jan, and Radek Škoda. "Polycrystalline Diamond Films as Protection of Zircaloy Fuel Cladding." In 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-30873.

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Polycrystalline diamond coating is a promising possibility for prevention, or reduction of high temperature oxidation of zirconium alloys. Zirconium alloys are used as cladding material in almost all types of nuclear reactors, where creates a barrier between nuclear fuel and cooling water in the primary circuit. Hydrogen and considerable amount of heat is released during steam oxidation that may occur in an eventual accident. In this paper Zircaloy-2 alloy was covered by polycrystalline diamond layer using Plasma Enhanced Linear Antennas Microwave Chemical Vapor Deposition system reactor. X-Ray Diffraction and Raman spectroscopy measurements confirmed coverage of the surface area with crystalline and amorphous carbon layer. Characterizations were done for zirconium alloy covered with diamond layer before and after corrosion and irradiation tests - ion beam irradiation tests and high temperature steam exposure.
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Srivastava, D., K. V. Mani Krishna, S. Neogy, G. K. Dey, I. Samajdar, and S. Banerjee. "Evolution of Microstructure, Microtexture and Texture in Dilute Zirconium Based Structural Components of Pressurised Heavy Water Reactors." In 17th International Conference on Nuclear Engineering. ASMEDC, 2009. http://dx.doi.org/10.1115/icone17-75274.

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Microstructure and texture of Zr alloys plays a significant role in deciding the material properties and its performance in nuclear reactor. In order to understand the development of microstructure, it is important to know the phase transformations mechanism in these alloys. Dilute Zr-Nb alloys exhibit a range of diffusional, diffusion less and hybrid phase transformations which includes martensitic omega and hydride transformation. With these points in view detail studies on the physical metallurgy aspects of the binary, ternary and quaternary Zirconium base alloys has been carried out in this study. Electron Diffraction and X-ray diffraction techniques have been used to determine the phases, defects and their crystallographic features. Hydriding is most important corrosion problem in the Zirconium cladding alloys. The mechanism of hydride phase transformation and their crystallographic and microstructural aspects has been presented in detail in some Zirconium alloys. Zirconium alloys being crystallographically asymmetric it shows deformation and annealing texture. Result of the studies on bulk and micro texture in some zirconium alloys and some micro-textural aspects on hydride formation is presented in this paper.
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Rebak, Raul B. "The Long-Term Environmental Degradation of Zirconium Alloys in Contact With Spent Nuclear Fuel: A Review." In ASME 2008 Pressure Vessels and Piping Conference. ASMEDC, 2008. http://dx.doi.org/10.1115/pvp2008-61272.

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Zirconium (Zr) alloys were mostly developed for nuclear power applications. The most common commercial alloys are Zircaloy, which are used as cladding for fuel pellets in water-cooled reactors. These alloys have adequate corrosion resistance in service under irradiation in presence of high-purity water at approximately 300–350°C but in a few cases they may suffer environmental degradation. The most common types of degradation are hydriding, shadow corrosion and nodular corrosion. After the fuel rod bundles are removed from the reactors they are temporarily stored in water pools until their permanent disposition nuclear waste repositories. Simulated laboratory testing and modeling show that the long-term storage in the sealed containers would not cause further damage to the cladding material until the waste containers are breached by corrosion many thousands of years later. And even after water incursion, since the temperature will be low, it is predicted that the cladding material would survive for many thousands of years more, thus delaying the release of the radionuclides to the atmosphere.
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Steinbru¨ck, Martin. "Oxidation of Advanced Zirconium Alloys in Oxygen in the Temperature Range 600-1600°C." In 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48054.

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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. Therefore, at FZK an experimental program on the high-temperature behavior of different cladding alloys presently used has been started. 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 one can say that the advanced cladding alloys here studied show also a favorable behavior at high temperatures during accident scenarios.
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Ding, Jie, Yixiong Zheng, Yang Ding, Song Liu, Libing Zhu, and Jiazheng Liu. "Test Reactor Irradiation Fuel Assembly Development in China." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-66617.

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During the development of zirconium alloys, the irradiation in the test reactor is a critical step to comparison the irradiation properties of candidate alloys, such as corrosion, creep and irradiation growth. In this paper, a small scaled fuel assembly for test reactor irradiation is designed, which meets the needs of new zirconium alloys development. The irradiation fuel assembly (IFA) can be easily disassembled, and the test fuel rods or irradiation specimen can be easily replaced, which makes it possible to do the further post-irradiation examination in the hot cell to obtain the irradiation performance data. Now the IFA has finish fabrication and the test reactor irradiation program is planned to launch in 2017.
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Booth, Martin, and Michael Martin. "Use of the Extended Finite Element Method in the Assessment of Delayed Hydride Cracking." In ASME 2016 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/pvp2016-63156.

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Zirconium alloys, as used in water-cooled nuclear reactors, are susceptible to a time-dependent failure mechanism known as Delayed Hydride Cracking, or DHC. Corrosion of zirconium alloy in the presence of water generates hydrogen that subsequently diffuses through the metallic structure in response to concentration, temperature and hydrostatic stress gradients. As such, regions of increased hydrogen concentration develop at stress concentrating features, leading to zirconium hydride precipitation. Regions containing zirconium hydride are brittle and prone to failure if plant transient loads are sufficient. This paper demonstrates the application of the Extended Finite Element Method, or XFEM, to the assessment of the DHC susceptibility of stress concentrating features, typical of those considered in the structural integrity assessment of heavy water pressure tube reactors. The method enables the calculation of a DHC threshold load. This paper builds on the process-zone approach that is currently used to provide the industry-standard DHC assessment of zirconium alloy pressure tubes and also recent developments that have extended the application of the process-zone approach to arbitrary geometries by the use of finite element cohesive-zone analysis. In the standard cohesive-zone approach, regions of cohesive elements are situated in discrete locations where the formation of zirconium hydride is anticipated. In contrast, the use of XFEM based cohesive formulations removes the requirement to define cohesive zones a priori, thereby allowing the assessment of geometries in which the location of hydride material is not known.
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Zhao, Zheng, Beibei Feng, Xingtuan Yang, and Yanfei Sun. "Prospect of MAO Technology Application in Nuclear Power Industry." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-15435.

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Micro arc oxidation (MAO) technology known as a newly surface treatment technology has got a widely application in the field of aviation, aerospace, automotive, electronics, and medical industry. Strength, toughness, hardness and corrosion of valve metal such as aluminum, magnesium, copper, zinc, zirconium and their alloys can be greatly improved by MAO technology. This paper tries to probe into the feasibility of using MAO technology in nuclear power industry. Aluminum and its alloys are used as structural materials such as the cladding of reactor fuel and all kinds of pipes in the low nuclear reactor. Zirconium alloys are widely used for the fuel cladding, cannula, catheter and other components of the fuel assemblies. Titanium and its alloys offer a unique combination of desirable mechanical properties which makes them to be the candidate materials for structural application in the field of nuclear energy. The surface of all these materials may be destroyed which increasing the risk of the nuclear accident due to the severe serving conditions. As a result, it is necessary to improve the corrosion and wear resistance behavior. With the urgent requirements of safety and durability of nuclear reactor, MAO technology must have a broad prospect in nuclear industry.
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Reports on the topic "Zirconium alloys – Corrosion"

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Johnson, A. B. Jr. Radiation effects on corrosion of zirconium alloys. Office of Scientific and Technical Information (OSTI), June 1989. http://dx.doi.org/10.2172/6316113.

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