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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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137 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Effect of Material Condition on Stress Corrosion Crack Initiation of Cold-Worked Alloy 600 in Simulated PWR Primary Water

The effect of material condition and grain boundary (GB) carbides on stress corrosion crack (SCC) initiation was evaluated on a cold-worked Alloy 600 heat in the solution annealed and thermally treated conditions. SCC initiation response of these material conditions were investigated by constant load testing in simulated pressurized water reactor primary coolant environments. Direct current potential drop technique was used for in-situ detection of SCC initiation. With periodic examination of the specimens, evolution of precursor damage including intergranular attack and short cracks was carefully documented and compared to data previously obtained from unstressed coupons with the same material conditions and were exposed to the same environment. Results revealed a predominant role of mechanical factors in the development of precursor damage with no apparent improvement of GB carbides on SCC initiation resistance.

Zhai, Ziqing↗
DOE OSTI2025

Time-Dependent 600 °C Post-Weld Heat Treatment Response of Laser-Wire-Directed Energy Deposited Austenitic Stainless Steel Claddings on Carbon Steel

Austenitic stainless steel overlay claddings are often used to protect nuclear reactor pressure vessels from corrosion. Arc welding processes are traditionally employed, which require a high heat input to bond the dissimilar metals, creating a large heat-affected zone and high residual stress that demands a lengthy stress-relieving post-weld heat treatment, up to 48 hours. High energy beam (e.g., laser, electron beam) directed energy deposition-based overlay cladding processes show promise for fabricating claddings with superior performance compared to arc welded ones because of their localized heat input; however, their necessity and time of post-weld heat treatment are currently unclear. In this research, we investigate the time dependence of a 600 C stress-relieving post-weld heat treatment of austenitic stainless steel clad on low carbon steel by laser-wire-directed energy deposition, comparing the microstructure and mechanical properties of the as-fabricated condition to 2-, 10-, and 48-hour heat treatment times. Diffusion simulations and microscopy show a gradual carburization of the stainless steel cladding and decarburization of the heat-affected zone/base metal. Nanoindentation testing shows up to a 1.8–2.69 embrittlement in the cladding and a 0.33 to 0.69 softening in the heat-affected zone after the 48-hour post-weld heat treatment. Tensile testing of the cladding composites reveals that increasing post-weld heat treatment time raises clad cracking susceptibility due to intergranular carbide formation and shows that laser-wire-directed energy deposition claddings may not require post-weld heat treatment.

Coating↗
DOE OSTI2019

Examination of Cavity Formation in Cold-Worked Alloy 690

Although stress corrosion cracking (SCC) has not been reported in Alloy 690 components after more than 30 years of pressure water reactor (PWR) service, susceptibility to SCC growth in PWR primary water has been reported by many labs for highly CW Alloy 690 along with isolated observations of crack initiation. Post-test examinations on SCC initiation specimens suggested a key role of creep by inducing the formation and growth of grain boundary cavities at carbide interfaces leading to intergranular cracking. However, it is clear that high levels of cold work can produce grain boundary voids and complicates the quantitative evaluation of damage evolution during testing in high-temperature water. Therefore, it is crucial to understand how to properly examine and classify this type of damage in the pre-test condition and establish its evolution with time. This paper outlines our observational protocols on how to best prepare sample surfaces to avoid inducing artifacts, and to perform subsequent examinations utilizing scanning electron microscopy, transmission electron microscopy and focused ion beam milling to assess the formation and development of grain boundary cavities. Detailed examinations will be reviewed on several commercial Alloy 690 heats in the pre-test CW condition and after long-term initiation testing at constant load in 360°C simulated PWR primary water to effectively document grain boundary cavity formation and growth leading to crack initiation.

Olszta, Matthew J.↗
DOE OSTI2019

Crack Initiation Behavior of Cold-Worked Alloy 690 in Simulated PWR Primary Water - Role of Starting Microstructure, Applied Stress and Cold Work on Precursor Damage Evolution

Long-term stress corrosion crack (SCC) initiation behavior in Alloy 690 is being evaluated by constant load tensile testing in simulated pressurized water reactor primary water. Seven commercial Alloy 690 heats with different starting microstructures and cold-work levels are being tested at either 90% or 100% of the material yield stress. Previous studies have revealed grain boundary (GB) cavity formation after 7,110 to 9,220 hours of exposure in certain thermally treated (TT) Alloy 690 heats in highly cold-worked (CW) conditions, serving as precursors to intergranular cracks. This paper summarizes the new characterization results for the evolution of GB cavities in four Alloy 690TT heats in 21% and 31%CW conditions loaded at yield stress after 18,400 hours of exposure. Link-up of GB cavities has led to formation of long internal creep cracks, resulting in crack initiation detected by direct current potential drop in a highly CW Alloy 690TT heat with uniform, semi-continuous distribution of small GB carbides. In comparison, highly CW materials featuring larger and more widely spaced GB carbides did not show a significant increase in GB cavity density or size. Quantitative assessments were performed on GB cavity evolution in highly CW specimens for two Alloy 690TT heats demonstrating the strong dependence of GB cavity evolution on applied stress and starting GB microstructure.

Zhai, Ziqing↗
DOE OSTI2020

Crack Growth Rate and Fracture Toughness Tests on Irradiated Ex-Plant Materials

The performance of structural materials is critical for the safe and economic operation of light water reactors. Exposed to neutron irradiation during service, the reactor core internal materials can undergo significant microstructural and microchemical changes, leading to irradiation hardening and embrittlement. To ensure the structural integrity and functionality of nuclear reactor components during long-term operation, material degradation and damage mechanisms must be understood and managed adequately. In this work, irradiated materials harvested from the decommissioned Zorita reactor were studied for their cracking susceptibility and fracture resistance as a function of irradiation dose up to 47 displacement per atom (dpa). The material is a Type 304 stainless steel sectioned from the baffle plates of this pressurized water reactor with 38 years of service. Crack growth rate and fracture toughness J-resistance (J-R) curve tests were performed in low-corrosion-potential environments at ~315°C. All samples behaved similarly under cyclic loading, and no deteriorated corrosion-fatigue behavior was observed in the test environments. Under constant stress intensity factors, most samples did not show elevated crack growth rates, suggesting an adequate stress corrosion cracking resistance in the test environments. However, an unstable cracking behavior was observed in a 47-dpa sample, resulting in significantly higher crack growth rates than expected at high stress intensity factors. Crack instability was also observed in a 0.06-dpa sample but did not lead to a sustained high crack growth rate. The impact of neutron irradiation was more evident in the fracture toughness J-R curve tests. As the dose increased, the J-R curve declined considerably and became very shallow at high doses. A fully intergranular fracture morphology was also observed among the high-dose samples ruptured in an air atmosphere at room temperature. This brittle fracture mode in the absence of high temperature water environment confirmed a high degree of embrittlement of this material resulting from its service exposure to neutron irradiation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

A study on tensile properties of Alloy 709 at various temperatures

In recent years, there have been several advancements in energy production from both fossil fuels and the alternate “clean” sources such as nuclear fission. These advancements are fueled by the need for more efficient systems that will optimize the use of the depleting fossil fuel reserves and shift the focus to cleaner sources of energy. The efficiency of any power generation cycle is dependent on the ability of the structural material to withstand increased peak operating temperatures. Advanced austenitic stainless steels have been in the focus as structural material for the next generation nuclear power plants, due to their strength, corrosion resistance, weldability and the wide range of temperatures at which the austenite phase is stable. Alloy 709, a recently developed advanced austenitic stainless steel, is being investigated in this paper. Here in this study, tensile tests were conducted on dog-bone samples of Alloy 709 in an in-situ scanning electron microscope (SEM) loading and heating stage, equipped with electron backscatter diffraction (EBSD), at various temperatures. The in-situ experiments indicated that the material primarily accommodated deformation by slip at lower temperatures. Void formation and coalescence at grain boundaries preceded slip at higher temperatures. Although crack initiation at all elevated temperatures was intergranular, the crack propagation through the material and the final fracture was transgranular ductile. Additionally, tensile tests were conducted on larger cylindrical samples at 550, 650 and 750 °C in air. The results of tests conducted in air and in-situ were found to be in agreement, at these temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Chrome-free qandilite (Mg 2 TiO 4 ) refractory aggregates: Role of titania source and evaluation of thermal expansion coefficient

As used chrome-based refractories may contain toxic Cr 6+ , the chrome-free Mg 2 TiO 4 phase is being considered which is known to improve hot strength and resistance to thermal shock and slag corrosion. As in situ Mg 2 TiO 4 formation generates porosity, preformed aggregates are desirable. Dilatometric studies revealed that when dead burned magnesia is combined with industrial anatase instead of rutile, it expanded less (1% versus 7.2%) and the reaction occurred earlier (965°C compared to 1120°C). After sintering at 1,600°C, rutile produced an aggregate with 3.8% open porosity, whereas anatase led to 0 % porosity and a bulk density of 3.24g/cm 3 with a thermal expansion coefficient (70–1,600°C) of 13.5×10 -6 K -1 . Moreover, replacing rutile by anatase eliminates the undesirable MgTiO 3 phase. The anatase-based aggregate consisted of a Mg 2 TiO 4 matrix with residual unreacted MgO and an intergranular, refractory CaTiO 3 phase. In conclusion, the rutile-based aggregate was deemed unsuitable due to a low-melting microstructure of forsterite contiguous with CaTiO 3 .

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Role of Cr-rich carbide precipitates in the intergranular oxidation of Ni-Cr alloys

The influence of grain boundary Cr carbides on the intergranular oxidation behavior of a Ni-16Cr alloy is considered using atomistic-to-mesoscale model and atom probe tomography. Cr carbide strongly perturbs the collective reactive dynamics of oxidizing species and alloy elements in the intergranular region. Strong attractive interactions between oxygen and carbide create a driving force for Cr and Ni accumulation in the grain boundary adjacent to the carbide, and for the depletion of Cr and Ni ahead of the oxidation front beyond the carbide. Furthermore, the results shed light on the mechanism of carbide-assisted protection of the alloy against grain boundary corrosion/oxidation.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Alternative PWHT to Improve High-Temperature Mechanical Properties of Advanced 9Cr Steel Welds

Creep-resistant 9Cr steels are extremely important in thermal power generation industry due to their marked resistance to creep and corrosion. The weldability of these alloys is critical since they are used in welded construction equipment. The required mechanical properties are achieved after post-weld heat treatment. This study examined the effect of different post-weld heat treatments on microstructure and mechanical properties of creep strength-enhanced 9Cr steel welding deposits. It was obtained with an experimental flux-cored arc welding wire used under protective gas (Ar-20% CO{sub 2}). The heat treatments used were: (1) tempering (760 °C × 2 h), (2) solubilizing (1050 °C × 1 h) + tempering (760 °C × 2 h) and (3) solubilizing (1150 °C × 1 h) + first tempering (660 °C × 3 h) + second tempering (660 °C × 3 h). All-weld metal chemical composition was analyzed, and hot tensile tests were carried out at different temperatures. Charpy-V impact tests and Vickers microhardness measurements were also performed. Microstructures were studied using x-ray diffraction and optical and scanning electron microscopy. In all cases, a martensitic matrix with intergranular and intra-granular precipitates was detected. In the as-welded condition, δ-ferrite was also found. Microhardness dropped, and the impact energy increased with post-weld heat treatments. The highest hot tensile strength result was achieved with samples submitted to austenization at 1150 °C and double tempering at 660 °C.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Identifying multiple synergistic factors on the susceptibility to stress relaxation cracking in variously heat-treated weldments

The 347H austenitic stainless steel has been widely used for pressure vessels and pipeline (PVP) applications due to its excellent creep and corrosion resistance, which fit ideally to the harsh conditions in petrochemical industries, fossil fuel or nuclear power plants, and modern energy storages. However, a failure mode has been commonly observed with cracks emerging at the heat affected zone (HAZ) of weldments during post-weld heat treatment (PWHT) or under intermediate to high temperature service conditions. This phenomenon is termed as Stress Relaxation Cracking (SRC) since the purpose of PWHT is to relieve the welding-induced residual stress fields, or as Stress Age Cracking (SAC) if failure happens during service. A leading literature explanation of this failure suggests that the residual stress relaxation and the precipitation dissolution and/or re-precipitation occur in the same temperature range, which can lead to locally high strains and thus to crack at the grain boundaries. Since in situ spatial measurements of residual stress fields, microstructural evolution, and failure processes are nearly infeasible, this work recourses to a micromechanical finite element framework that models the high temperature failure as the nucleation and growth of grain boundary cavities, whereas various parameters such as thermomechanical loading history and its evolution, the competition of grain-interior dislocation creep and grain-boundary diffusion in failure lifetime, and microstructural heterogeneities (such as the precipitate free zone near grain boundaries) can be quantitatively incorporated. It can be concluded from these microstructure-explicit simulations that an accurate knowledge of residual stress evolution and a carefully calibrated set of material constitutive parameters are the essential prerequisites for lifetime predictions. The understanding of individual governing factors also leads to a mechanistic interpretation of the observed SRC susceptibility C-curves. In conclusion, these results suggest that the criticality of residual stress evolution, but not the precipitation-induced local strains, be the leading factor for SRC.

347H stainless steel weldments↗