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Featured researches published by George A. Young.


Metallurgical and Materials Transactions A-physical Metallurgy and Materials Science | 1995

Aqueous environmental crack propagation in high-strength beta titanium alloys

Lisa M. Young; George A. Young; John R. Scully; Richard P. Gangloff

The aqueous environment-assisted cracking (EAC) behavior of two peak-aged beta-titanium alloys was characterized with a fracture mechanics method. Beta-21S is susceptible to EAC under rising load in neutral 3.5 pct NaCl at 25 °C and −600 mVSCE, as indicated by a reduced threshold for subcritical crack growth (KTH), an average crack growth rate of up to 10 μms, and intergranular fracture compared to microvoid rupture in air. In contrast, the initiation fracture toughness (KICi) of Ti-15-3 in moist air is lower than that of Beta-21S at similar high σYS (1300 MPa) but is not degraded by chloride, and cracking is by transgranular microvoid formation. The intergranular EAC susceptibility of Beta-21S correlates with both α-colonies precipitated at β grain boundaries and intense slip localization; however, the causal factor is not defined. Data suggest that both features, and EAC, are promoted by prolonged solution treatment at high temperature. In a hydrogen environment embrittlement (HEE) scenario, crack-tip H could be transported by planar slip bands to strongly binding trap sites and stress/strain concentrations at α colony or β grain boundaries. The EAC in Beta-21S is eliminated by cathodic polarization (to −1000 mVSCE), as well as by static loading for times that otherwise produce rising-load EAC. These beneficial effects could relate to reduced H production at the occluded crack tip during cathodic polarization and to increased crack-tip passive film stability or reduced dislocation transport during deformation at slow crack-tip strain rates. High-strength β-titanium alloys are resistant, but not intrinsically immune to chloride EAC, with processing condition possibly governing fracture.


Other Information: PBD: 5 Apr 2003 | 2003

The Stress Corrosion Crack Growth Rate of Alloy 600 Heat Affected Zones Exposed to High Purity Water

George A. Young; Nathan Lewis

Grain boundary chromium carbides improve the resistance of nickel based alloys to primary water stress corrosion cracking (PWSCC). However, in weld heat affected zones (HAZs), thermal cycles from fusion welding can solutionize beneficial grain boundary carbides, produce locally high residual stresses and strains, and promote PWSCC. The present research investigates the crack growth rate of an A600 HAZ as a function of test temperature. The A600 HAZ was fabricated by building up a gas-tungsten-arc-weld deposit of EN82H filler metal onto a mill-annealed A600 plate. Fracture mechanics based, stress corrosion crack growth rate testing was performed in high purity water between 600 F and 680 F at an initial stress intensity factor of 40 ksi {radical}in and at a constant electrochemical potential. The HAZ samples exhibited significant SCC, entirely within the HAZ at all temperatures tested. While the HAZ samples showed the same temperature dependence for SCC as the base material (HAZ: 29.8 {+-} 11.2{sub 95%} kcal/mol vs A600 Base: 35.3 {+-} 2.58{sub 95%} kcal/mol), the crack growth rates were {approx} 30X faster than the A600 base material tested at the same conditions. The increased crack growth rates of the HAZ is attributed to fewer intergranular chromium rich carbides and to increased plastic strain in the HAZ as compared to the unaffected base material.


Corrosion | 2000

The Influence of Dissolved hydrogen on Nickel Alloy SCC: A Window to Fundamental Insight

David S. Morton; Steven A. Attanasio; George A. Young; Peter L. Andresen; Thomas M. Angeliu

Prior stress corrosion crack growth rate (SCCGR) testing of nickel alloys as a function of the aqueous hydrogen concentration (i.e., the concentration of hydrogen dissolved in the water) has identified different functionalities at 338 and 360 C. These SCCGR dependencies have been uniquely explained in terms of the stability of nickel oxide. The present work evaluates whether the influence of aqueous hydrogen concentration on SCCGR is fundamentally due to effects on hydrogen absorption and/or corrosion kinetics. Hydrogen permeation tests were conducted to measure hydrogen pickup in and transport through the metal. Repassivation tests were performed in an attempt to quantify the corrosion kinetics. The aqueous hydrogen concentration dependency of these fundamental parameters (hydrogen permeation, repassivation) has been used to qualitatively evaluate the film-rupture/oxidation (FRO) and hydrogen assisted cracking (HAC) SCC mechanisms. This paper discusses the conditions that must be imposed upon these mechanisms to describe the known nickel alloy SCCGR aqueous hydrogen concentration functionality. Specifically, the buildup of hydrogen within Alloy 600 (measured through permeability) does not exhibit the same functionality as SCC with respect to the aqueous hydrogen concentration. This result implies that if HAC is the dominant SCC mechanism, then corrosion at isolated active path regions (i.e., surface initiation sites or cracks) must be the source of localized elevated detrimental hydrogen. Repassivation tests showed little temperature sensitivity over the range of 204 to 360 C. This result implies that for either the FRO or the HAC mechanism, corrosion processes (e.g., at a crack tip, in the crack wake, or on surfaces external to the crack) cannot by themselves explain the strong temperature dependence of nickel alloy SCC.


Reference Module in Materials Science and Materials Engineering#R##N#Comprehensive Nuclear Materials | 2012

Welds for Nuclear Systems

George A. Young; M.J. Hackett; Julie D. Tucker; T.E. Capobianco

Nuclear power systems are constructed from a wide range of metallic alloys subjected to taxing environmental conditions and required to resist cracking and degradation of their principal mechanical and physical properties for decades. Fusion welding is, in general, the joining method of choice because of its hermeticity, high joint efficiency, and economic advantages relative to mechanical or brazed joints. However, it is often fusion welds or their heat-affected zones that prematurely degrade or fail because of the complex interplay of physical defects, compositional gradients, metallurgical changes, and residual stresses. This chapter presents the current mechanistic understanding of welding defects, reviews recent developments in assessing residual stresses and plastic strains, and relates these factors to the in-service performance of welds. Finally, the weldability of common structural alloy systems is reviewed.


15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors | 2011

Physical Metallurgy, Weldability, and in-Service Performance of Nickel-Chromium Filler Metals Used in Nuclear Power Systems

George A. Young; Robert A. Etien; Micah J. Hackett; Julie D. Tucker; Thomas E. Capobianco

Wrought Alloy 690 is well established for corrosion resistant nuclear applications but development continues to improve the weldability of a filler metal that retains the corrosion resistance and phase stability of the base metal. High alloy Ni-Cr filler metals are prone to several types of welding defects and new alloys are emerging for commercial use. This paper uses experimental and computational methods to illustrate key differences among welding consumables. Results show that solidification segregation is critical to understanding the weldability and environmentally-assisted cracking resistance of these alloys. Primary water stress corrosion cracking tests show a marked decrease in crack growth rates near 21 wt. % Cr at the grain boundary. While filler metals with 21–29 wt.% grain boundary Cr show similar PWSCC resistance, the higher alloyed grades are more prone to solidification cracking. Modeling and aging studies indicate that in some filler metals minor phase formation (e.g., Laves and σ) and long range order (LRO) must be assessed to ensure adequate weldability and inservice performance.


Environmental Degradation of Materials in Nuclear Power Systems | 2017

Influence of Alloying on α-αʹ Phase Separation in Duplex Stainless Steels

David A. Garfinkel; Jonathan D. Poplawsky; Wei Guo; George A. Young; Julie D. Tucker

Thermal embrittlement caused by phase transformations in the temperature range of 204–538 °C limits the service temperature of duplex stainless steels. The present study investigates a set of wrought (2003, 2101, and 2205) and weld (2209-w and 2101-w) alloys in order to better understand how alloying elements affect thermal embrittlement. Samples were aged at 427 °C for up to 10,000 h. The embrittlement and thermal instability were assessed via nanoindentation, impact toughness testing, and atom probe tomography (APT). Results demonstrate that the spinodal amplitude is not an accurate predictor of mechanical degradation, and that nanoindentation within the ferrite grains served as a reasonable approximate for the embrittlement behavior. Compositionally, alloys with a lower concentration of Cr, Mo, and Ni were found to exhibit superior mechanical properties following aging.


Metallurgical and Materials Transactions A-physical Metallurgy and Materials Science | 2002

The effects of test temperature, temper, and alloyed copper on the hydrogen-controlled crack growth rate of an Al-Zn-Mg-(Cu) alloy

George A. Young; John R. Scully


Acta Materialia | 2015

Assessment of thermal embrittlement in duplex stainless steels 2003 and 2205 for nuclear power applications

Julie D. Tucker; M.K. Miller; George A. Young


Nanotechnology | 2016

An atom probe perspective on phase separation and precipitation in duplex stainless steels

Wei Guo; David A. Garfinkel; Julie D. Tucker; Daniel Haley; George A. Young; Jonathan D. Poplawsky


JOM | 2012

Effect of Travel Speed and Beam Focus on Porosity in Alloy 690 Laser Welds

Julie D. Tucker; Terrance K. Nolan; Anthony J. Martin; George A. Young

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Jonathan D. Poplawsky

Oak Ridge National Laboratory

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Wei Guo

Oak Ridge National Laboratory

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M.K. Miller

Oak Ridge National Laboratory

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