Microstructure and corrosion behavior of Zr-702 joined by electron beam welding

Zhang, Binggang; Li, Xiao Peng; Wang, Ting; Wang, Xudong

doi:10.1016/j.vacuum.2015.08.005

Cited by 34 publications

(9 citation statements)

References 15 publications

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“…The elemental analysis by EDS technique showed uniform distribution of Zr and Nb over welding region without any Nb precipitations. Similar microstructure behavior of Zr-Nb and Zr-Sn alloys were also found after EBW welding in [29]. Most likely that the temperature during welding process was enough for dissolution any precipitates in zirconium matrix.…”

Section: Sample Characterizationsupporting

confidence: 67%

High-Temperature Oxidation of Cr-Coated Resistance Upset Welds Made from E110 Alloy

2021

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The resistance upset welds (RUW) made from E110 alloy without and with Cr coatings were oxidized in air atmosphere at 1100 °C for 2, 10 and 30 min. The cross-section microstructure, elemental composition and hardness were studied before and after oxidation using optical and scanning electron microscopy, and indentations in welding region. The RUW welding does not noticeably change oxidation kinetics of E110 alloy. The most crucial effect has surface non-regularities formed after welding, which prevent uniform coating deposition on full surface of welded cladding tube and end plug. Cr coating deposition can strongly reduce oxidation of welded E110 alloy, while additional post-processing treatment should be applied to improve surface morphology after RUW welding. Several suggestions favorable to development of ATF Zr-based claddings using Cr coating deposition on welded nuclear rods were discussed.

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Section: Sample Characterizationsupporting

confidence: 67%

High-Temperature Oxidation of Cr-Coated Resistance Upset Welds Made from E110 Alloy

2021

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“…Semenov et al (2013) discovered that by optimizing the EBW heat source, the corrosion resistance of weld joints between Zr-2.5Nb and steel using a Zr-1Nb filler material is slightly improved when compared to complicated traditional post-weld thermomechanical treatments. Zhang et al (2015) conclude that EBW of Zr-702 actually improved the corrosion resistance of the joint region because less pitting was observed at the weld when compared to the heat-affected zone (HAZ) and base metal. Suzuki et al (1962) found that EBW of Zr and Zr-2 sheet produces weld joints with better mechanical properties and corrosion resistance in high-temperature-pressure water than gas tungsten arc-welded specimens.…”

Section: Introductionmentioning

confidence: 86%

“…These fuels are defined as nuclear fuel systems that can outperform current fuel systems (i.e., standard UO 2 -Zr-based alloy light water reactor fuel assemblies) during accident conditions (such as a loss-of-coolant accident) and maintain similar or higher performance during normal reactor operations, operational transients, and design-basis and beyond design-basis accidents. Zinkle et al (2014) describes the three main strategies being pursued in regards to accident-tolerant fuels: (1) improvement of oxidation resistance of current zirconium alloy cladding, (2) replacement of zirconium alloys with alternative oxidationresistant high-performance cladding, and (3) replacement of monolithic ceramic oxide fuel with alternative fuel forms. Parga (2013) provides an extensive analysis on the aggressive environment experienced by fuel rods during a nuclear reactor severe accident based on the analysis of actual reactor core meltdowns (e.g., Three Mile Island-2, Chernobyl Reactor-4, and Fukushima Daiichi) and nuclear reactor safety tests.…”

Section: Introductionmentioning

confidence: 99%

Room temperature mechanical properties of electron beam welded zircaloy-4 sheet

Parga

Rooyen

Coryell

et al. 2017

Journal of Materials Processing Technology

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Resumption of operations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory was approved in 2014 to meet U.S. Department of Energy Office of Nuclear Energy objectives in transient testing of nuclear fuels. In parallel, the National Nuclear Security Administration, through the Office of Material Management and Minimization, is converting TREAT from its existing highly enriched uranium core to a low-enriched uranium core. This effort entails designing, fabricating, and qualifying a new TREAT low-enriched uranium fuel assembly, while maintaining TREAT's experimental performance capabilities. Zircaloy-4 is being evaluated as TREAT lowenriched uranium assembly material. A preliminary study on the room temperature mechanical properties of asreceived and electron beam welded Zircaloy-4 sheet (1.6 mm) is presented. The sheet was high-vacuum electron beam welded using a three-pass process with varying heat input: (1) tack welding (10.5 J/mm), (2) seams welded (36.7 J/mm), and (3) sealed (15.7 J/mm). As-received and electron beam welded specimens show comparable properties. Zircaloy-4 displays anisotropy between the transverse and longitudinal directions. Tensile properties measured for the transverse direction display higher yield strength, reduction of area, and slightly lower tensile strength and ductility than for the longitudinal (i.e., rolling) direction. Weld and base metal hardness are comparable, while hardness at the heat-affected-zone is slightly higher. Microscopic examinations show distinct microstructure morphology and grain size from weld to base metal. A correlation between welding parameters, mechanical properties, and microstructural features was established for electron beam welded Zircaloy-4 sheet.

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“…Improving the stress corrosion resistance and structural strength of post-treatment zirconium alloys has always been a focus of attention for scholars both domestically and abroad. Therefore, the study of SCC has become an important issue in the field of Zr alloy applications [11]. Bernard et al [12] studied the corrosion and SCC properties of zirconium used in the La Hague nuclear fuel reprocessing plant in France and reported that zirconium is susceptible to SCC, as indicated by a remarkable ductility loss, at the passivation breakdown potential.…”

Section: Introductionmentioning

confidence: 99%

Strain rate susceptibility of stress corrosion cracking for commercial Zr702 used in spent nuclear fuel reprocessing

Liu

et al. 2023

Corrosion Engineering, Science and Technology

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Slow strain rate tensile (SSRT) experiments were applied to comprehensively examine the strain rate susceptibility of the stress corrosion cracking behaviour of commercial Zr702 used in the spent nuclear fuel reprocessing environment. Results revealed that layers formed on Zr702 in boiling HNO 3 solutions was composed of monoclinic ZrO 2 (m-ZrO 2 ) and tetragonal ZrO 2 (t-ZrO 2 ). With strain rates above 10 −4 s −1 , oxide layers ruptured rapidly, evidenced by the drop down of opened circuit potential (OCP). However, the OCP curve displayed fluctuation with 10 −5 s −1 during straining, indicating that oxide layer alternately ruptured and repaired, resulting in the thickening of the oxide layer and increasement of stress corrosion cracking susceptibility.

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Microstructure and corrosion behavior of Zr-702 joined by electron beam welding

Cited by 34 publications

References 15 publications

High-Temperature Oxidation of Cr-Coated Resistance Upset Welds Made from E110 Alloy

High-Temperature Oxidation of Cr-Coated Resistance Upset Welds Made from E110 Alloy

Room temperature mechanical properties of electron beam welded zircaloy-4 sheet

Strain rate susceptibility of stress corrosion cracking for commercial Zr702 used in spent nuclear fuel reprocessing

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