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The Precipitation and Redistribution of Alloying Element in Zircaloy-4 Cladding Tube Oxidized in High-Temperature Steam
Zirconium (Zr)-based alloys are widely used as fuel cladding material for light water reactors. Under a loss-of-coolant accident (LOCA) condition,...
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Flow Stress Modeling of Tube and Slab Route Sheets of Zircaloy-4 Using Machine Learning Techniques and Arrhenius Type Constitutive Equations
The present work examines the effectiveness of a strain-compensated Arrhenius-type constitutive model, Artificial Neural Network (ANN) and Support...
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Improved oxidation resistance of Cr-Si coated Zircaloy with an in-situ formed Zr2Si diffusion barrier
In the present study, the dense Cr and Cr 0.92 Si 0.08 coatings have been deposited on Zircaloy-4 substrates by the magnetron sputtering technique. The...
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Effect of Bias Voltage on the Microstructure, Mechanical Properties, and High-Temperature Steam Oxidation Behavior of Cr Coatings Prepared by Magnetron Sputtering on Zircaloy-4 Alloy
This study investigated the effect of bias voltage on the microstructure, mechanical properties, and high-temperature steam oxidation behavior of the...
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Correlations of the Steam Oxidation Rate Constant of BWR Alloy Zircaloy-2 at 800–1400 °C
Steam oxidation experiments were conducted at 800–1400 °C with boiling water reactor alloy Zircaloy-2 strip specimens. Sample weight gain...
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Effect of heat input on microstructure, mechanical and corrosion properties of electron beam welded zircaloy-4 sheets
The present study reports the effect of heat input on the evolution of microstructure, mechanical properties, and corrosion behavior of electron beam...
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Modeling Zirconium Alloys Recrystallization by Full-Field and Mean-Field Approaches
The present article details the results obtained with recrystallization models based upon mean-field or full-field assumptions, respectively. The two...
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Texture Analysis and Fracture Behavior of Zircaloy-4 Processed Through Swaging
Zirconium alloy (Zircaloy-4) was cold-worked using the swaging process and subsequently subjected to mechanical and microstructural characterization...
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Effect of Hydrogen Isotopes on Delayed Hydride Cracking Behavior of Zr-2.5Nb Pressure Tube Material
To study the effect of hydrogen isotopes on the delayed hydride cracking (DHC) behavior, Zr-2.5Nb pressure tube (PT) material was charged with about...
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Effect of Processing Routes on Orientation-Dependent Tensile Flow Behavior of Zircaloy-4 at Elevated Temperatures
The present work describes the correlation between orientation-dependent tensile flow behavior and texture of Zircaloy-4 sheet at elevated...
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Nuclear Fuel Complex—A Five Decade Success Story of Indian Nuclear Power Program
Nuclear Fuel Complex (NFC), Hyderabad was conceived as a pivotal industrial arm of the Department of Atomic Energy (DAE) in the late 60s with a...
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Low-energy electron-driven observation of nanometer-sized Laves phases at alloy surfaces enabling statistical characterization with high precision and efficiency
Alloys strengthened by nanometer-sized Laves phases have been used as structural components working in corrosive environments at high temperatures...
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Effect of hydrogen isotopes on tensile and fracture properties of Zr–2.5Nb pressure tube material
AbstractThe majority of data reported in open literature on hydride embrittlement of Zr alloys is based on hydrogen not deuterium which evolves...
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Effect of Process Variables on Texture and Hydride Orientation of Cold Pilgered Zr-Sn-Nb-Fe Cladding Tubes
Cold pilgering technology is one of the most widely used methods to fabricate the seamless zirconium alloy cladding tubes. In practice, the process...
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Enhanced oxidation resistance and adhesion performance of Cr–Mo-coated Zry-4 by a thin Cr layer in steam up to 1600 °C
In this study, Cr–Mo–Cr tri-layer coating with Cr inner layer is designed and compared with the pure Cr coating and Cr–Mo bi-layer coating in steam...
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Heat Resistance of Zr–1Nb Alloy Claddings in Water Vapor after Ion-Plasma Nitriding
The influence of ion-plasma nitriding time of Zr–1Nb alloy fuel cladding tubes on their resistance to oxidation in water vapor at 600–1200°C was...
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Creep Properties of 9Cr and 14Cr ODS Tubes Tested by Inner Gas Pressure
Oxide-dispersion strengthened steels are promising materials for extreme service conditions including nuclear reactors core. In service conditions,...
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Modeling of structural failure of Zircaloy claddings induced by multiple hydride cracks
Zirconium alloys have been serving as primary structural materials for nuclear fuel claddings. Structural failure analysis under extreme conditions...
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Polycrystalline Diamond and Cr Double Coatings Protect Zr Nuclear Fuel Tubes Against Accidental Temperature Corrosion in Water-Cooled Nuclear Reactors
The chapter proposes a new strategy for protecting Zr alloy nuclear fuel tubes from accidental temperature corrosion in water-cooled nuclear...
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Rapid assessment of structural and compositional changes during early stages of zirconium alloy oxidation
A multimodal chemical imaging approach has been developed and applied to detail the dynamic, atomic-scale changes associated with oxidation of a...
