Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “STRESS CORROSION”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 991 records · Page 55
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Effects of Co contents on the microstructures and properties of the electrodeposited NiCo–Zr composite coatings

Highlights: • The novel NiCo–Zr coatings were prepared by electro-deposition. • Surface morphology, crystal structure, grain size and microstrain were examined. • Texture, residual stress and corrosion resistance were investigated. • Addition of Co increased the hardness and corrosion resistance of the coatings. - Abstract: In this study, the NiCo–Zr composite coatings were prepared from the electrolytes with different Co{sup 2+} concentrations by electrodeposition method. The effects of Co contents on the crystal structure, surface morphology, grain size, microstrain and residual stress were examined by X-ray diffractometer (XRD), field emission scanning electron microscopy (FESEM), Energy dispersive X-ray spectroscopy (EDX) and atomic force microscope (AFM). The corrosion resistance of the composite coatings was also examined by the potentiodynamic polarization and electrochemical impedance (EIS) measurements. The results revealed that the crystal structures of the coatings were dependent on the Co contents and addition of Co content of 58 wt% resulted in the formation of hexagonal (hcp) Co. The increasing Co contents in the NiCo–Zr composite coatings resulted in the smoother and more compact surface, decreased the grain size and increased the microstrain. The micro-hardness and residual stress also increased with increasing Co contents. The addition of Co increased the corrosion resistance of the NiCo–Zr composite coatings compared with the Ni–Zr coating while the corrosion resistance of the NiCo–Zr composite coatings decreased as the Co contents increased.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Corrosion test of LWR cladding tubes in high temperature seawater

In the case of cooling system failure in spent fuel pools (SFP) due to station black out and/or loss of coolant accidents, a pool water level must be kept high enough to avoid overheat of spent fuels. Seawater is one of the candidate coolant materials in such emergencies. In the present work, we have studied the corrosion behavior of zirconium alloys in seawater. Tube samples of Zircaloy-2 and -4 claddings were immersed in artificial seawater (3, 10 and 20 wt.%) at high temperature (80 - 85 Celsius degrees). Boric acid was added to the seawater at a corrosion test of Zircaloy-4. Two types of specimens were prepared according to the simulation of circumferential tensile stress. One was a plugged tube specimen, which was pressurized to around fuel rod pressure in a spent fuel pool by argon gas. Another specimen was of C-ring shape, where large circumferential tensile stress was partially induced at the surface. A notch was longitudinally grooved to increase the local tensile stress on some of the C-ring specimens. The specimen surface was observed by a scanning electron microscopy (SEM) after the corrosion test. After the 200 day long corrosion test for the plugged tube specimens without boric acid, Mg(OH){sub 2} deposits were observed on the surface of Zircaloy-2 and -4 specimens whereas Mg(OH){sub 2} was not observed with boric acid because of probably a lower pH. Neither pitting corrosion nor stress corrosion cracking (SCC) was observed on the surface of all plugged tube specimens. Only in the C-ring specimens in 20 wt.% seawater without boric acid and in 3 and 10 wt.% seawater with boric acid, small cracks were observed at the notch. The results of the present work show that Zircaloy-2 and -4 have a high tolerance towards corrosion and SCC in high temperature seawater.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Quantifying Optical Loss of High-Voltage Degradation Modes in PV Modules Using Spectral Analysis [Slides]

The direct current bias for photovoltaic (PV) modules interconnected in series-strings may include both high voltage negative ("HV-") and positive ("HV+") polarity with respect to the electrical ground. Multiple degradation modes, resulting in quantifiable optical loss, were found to occur during HV-/HV+ sequential stress, including: corrosion of the external glass surface, encapsulant delamination (at its interfaces with the glass and the PV cell), internal haze formation (resulting from a chemical interaction between the glass and the encapsulant), corrosion and migration of the gridlines, and corrosion of the silicon nitride (SixNy) antireflective coating on the cell. The effects of these separate modes were examined using monolithic (e.g., glass or PV cell) and laminated-coupon (glass/encapsulant/glass or glass/encapsulant/cell/encapsulant/backsheet) specimens. Characterizations during and after unbiased accelerated testing at 85 degrees C/85% relative humidity included: spectrophotometry, optical microscopy, electron microscopy, and ellipsometry. For some module components (i.e., the glass and the SixNy coating), the optical performance was determined through iterative analysis of empirical measurements. Concentrating on just their spectral effect, a novel model was then developed to estimate the transfer of light to the PV cell and the return of light from the PV module with simultaneous degradation mechanisms, which was compared to a mini-module previously subjected to HV-/HV+ stress. The model suggests one-third of the current loss observed for the mini-module can be attributed to the optical degradation of the packaging materials. The dominant degradation modes include encapsulant delamination and corrosion of the SixNy coating. Recommendations are given so that the optical model may be improved relative to accelerated testing and validated relative to field aging.

14 SOLAR ENERGY↗
DOE OSTI2026

Flaw Tolerance Assessment for DOE Standard SNF Dry Storage Canisters - 26550

The U.S. DOE has designed four spent nuclear fuel (SNF) dry storage canisters for storing DOE standardized SNFs. The DOE standard canisters are cylindrical shells with a diameter of 24 inches (610 m) or 18 inches (457 m), a wall thickness of 0.5 inches (12.7 m) or 0.375 inches (9.53 m), and a length of 15 feet (4.57 m) or 10 feet (3.05 m). These DOE canister geometries are completely different from commercial canisters. The latter may experience chloride-induced stress cracking corrosion (CI-SCC) because they are stored near coastal regions. The former may not experience CI-SCC but face different challenges because they are stored in the SNF storage facilities. Because of large residual stresses, mechanical flaws may occur in the DOE canisters during long-distance transportation or lifting handling. To date, only limited structural integrity analyses were carried out through drop tests on the DOE canisters, but a more general flaw tolerance assessment has not been performed. Therefore, the failure assessment diagram (FAD)-based fracture mechanics method, as codified by the latest API 579-1/ASME FFS-1-2021 Edition, is adopted in this work to assess surface flaw tolerance for DOE canisters under operation loading and welding residual stresses (WRS), where the new code-recommended WRS distributions are used. To more adequately consider the transverse distribution of WRS, an equivalent WRS distribution is proposed to account for the WRS reduction with distance from the weld centerline. Moreover, the closed-form solutions of stress intensity factor K, which serves as the crack driving force during subcritical crack growth, are developed from the tabular data of the K factors provided in API 579-1/ASME FFS-1 and used to determine more accurate flaw sizes at flaw instability. Subsequently, the Level 2 assessment procedures with 12 assessment steps, as codified and detailed in API 579-1 and ASME FFS-1, are followed to assess the flaw tolerance for the surface flaws in the DOE standard canisters with consideration of normal or accident operation loads combined with WRS. The assessment results show that the four designs of DOE standard canisters can tolerate all surface flaws that meet the code permitted maximum sizes of a flaw length of 8 inches (i.e., 200 mm) and a flaw depth of 80% wall thickness. This demonstrates that all designs of DOE standard canisters are robust and reliable.

DOE standard canister↗
DOE OSTI2012

Improved cladding nano-structured materials with self-repairing capabilities

When designing nuclear reactors or the materials that go into them, one of the key challenges is finding materials that can withstand an outrageously extreme environment. In addition to constant bombardment by radiation, reactor materials may be subjected to extremes in temperature, physical stress, and corrosive conditions. A limitation in fuel burnup is and usage of the nuclear fuel material is related to the structural material radiation damage, that makes the fuel be removed with low-burnup and immobilized in the waste storage pools. The advanced burnup brings cladding material embitterment due to radiation damage effects corroborated with corrosion effects makes the fuel pellet life shorter. The novel nano-clustered structured sintered material may mitigate simultaneously the radiation damage and corrosion effects driving to more robust structural materials that may make the nuclear reactor safer and more reliable. The development of nano-clustered sinter alloys provides new avenues for further examination of the role of grain boundaries and engineered material interfaces in self-healing of radiation-induced defects driving to the design of highly radiation-tolerant materials for the next generation of nuclear energy applications. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Quantifying optical loss of high-voltage degradation modes in photovoltaic modules using spectral analysis

The direct current bias for photovoltaic (PV) modules interconnected in series-strings may include both high voltage negative (“HV-”) and positive (“HV+”) polarity with respect to the electrical ground. Multiple degradation modes, resulting in quantifiable optical loss, were found to occur during HV-/HV+ sequential stress, including corrosion of the external glass surface, encapsulant delamination (at its interfaces with the glass and the PV cell), internal haze formation (resulting from a chemical interaction between the glass and the encapsulant), corrosion and migration of the gridlines, and corrosion of the silicon nitride (Si x N y ) antireflective coating on the cell. The effects of these separate modes were examined using monolithic (e.g., glass or PV cell) and laminated-coupon (glass/encapsulant/glass or glass/encapsulant/cell/encapsulant/backsheet) specimens. Characterizations during and after unbiased accelerated testing at 85°C/85% relative humidity included spectrophotometry, optical microscopy, electron microscopy, and ellipsometry. For some module components (i.e., the glass and the Si x N y coating), the optical performance was determined through iterative analysis of empirical measurements. Concentrating on just their spectral effect, a novel model was then developed to estimate the transfer of light to the PV cell and the return of light from the PV module with simultaneous degradation mechanisms, which was compared with a mini-module previously subjected to HV-/HV+ stress. Here the model suggests that one third of the current loss observed for the mini-module can be attributed to the optical degradation of the packaging materials. The dominant degradation modes include encapsulant delamination and corrosion of the Si x N y coating. Recommendations are given so that the optical model may be improved relative to accelerated testing and validated relative to field aging.

14 SOLAR ENERGY↗
DOE OSTI2022

Correlating the Morphological Changes to Electrochemical Performance During Carbon Corrosion in Polymer Electrolyte Fuel Cells

A mechanistic understanding of carbon corrosion in polymer electrolyte fuel cells (PEFCs) is required to design durable catalyst layers. Uncontrolled startup and shutdown of PEFCs cause electrochemical oxidation of carbon, which leads to several degradation phenomena, such as loss in electrochemical surface area (ECSA), pore structure collapse or increase in mass transport resistance. In this study, the chronology of morphological changes in the cathode catalyst layer due to carbon corrosion was identified and correlated with electrochemical performance degradation. PEFCs were subjected to the Department of Energy carbon corrosion accelerated stress test (AST) protocol. The study revealed two phases: in the initial phase (~500 AST cycles), amorphous carbon in contact with Pt nanoparticles oxidized fast. Rapid carbon loss and catalyst layer thinning occurred, but pore structure did not change significantly. Pt nanoparticles detached from the support and ECSA decreased drastically. In the second phase (~1500 AST cycles), carbon corrosion slowed down, but severe pore structure collapse was observed. Porosity and pore connectivity within the cathode catalyst layer decreased considerably. Electrochemical diagnostics corroborated this finding by showing significantly higher O2 mass transport resistance. Lastly, no significant change was observed in the concentration of oxides on the carbon surface after AST. But overall water management in the cathode catalyst layer deteriorated as the pore structure collapsed. This study provides an in-depth understanding of morphological changes during PEFC carbon corrosion AST protocol and motivates novel material design strategies to enable durable PEFCs.

catalyst layers↗
DOE OSTI2017

Characterization of Al-Mg Alloy Aged at Low Temperatures

For this research, long-term aged [343 K (70 °C) for 30 months and natural exposure for over 10 years] Al 5456 H116 samples were characterized using electron backscatter diffraction (EBSD), scanning transmission electron microscopy (STEM), state-of-the-art energy-dispersive X-ray spectroscopy (EDS) systems, and small-angle neutron scattering (SANS). ASTM G-67 mass loss tests of the sensitized Al 5456 alloy samples were conducted. Intragranular Mg-rich precipitates, such as Guinier–Preston (GP) zones, were confirmed in Al 5456 H116 aged at 343 K (70 °C) for 30 months, and the volume of these precipitates is 1.39 pct. β' phase is identified at the grain boundary of a navy ship sample, while high-resolution STEM results reveal no intragranular precipitates. Intergranular corrosion (IGC) of Al 5456 was found to be related to the continuity of intergranular precipitates.

36 MATERIALS SCIENCE↗
NASA NTRS1972

Ion and laser microprobes applied to the measurement of corrosion produced hydrogen on a microscopic scale.

Use of an ion microprobe and a laser microprobe to measure concentrations of corrosion-produced hydrogen on a microscopic scale. Hydrogen concentrations of several thousand ppm were measured by both analytical techniques below corroded and fracture surfaces of hot salt stress corroded titanium alloy specimens. This extremely high concentration compares with only about 100 ppm hydrogen determined by standard vacuum fusion chemical analyses of bulk samples. Both the ion and laser microprobes were used to measure hydrogen concentration profiles in stepped intervals to substantial depths below the original corroded and fracture surfaces. For the ion microprobe, the area of local analysis was 22 microns in diameter and for the laser microprobe, the area of local analysis was about 300 microns in diameter. The segregation of hydrogen below fracture surfaces supports a previously proposed theory that corrosion-produced hydrogen is responsible for hot salt stress corrosion embrittlement and cracking of titanium alloys. These advanced analytical techniques suggest great potential for many areas of stress corrosion and hydrogen embrittlement research, quality control, and field inspection of corrosion problems. For example, it appears possible that a contour map of hydrogen distribution at notch roots and crack tips could be quantitatively determined. Such information would be useful in substantiating current theories of stress corrosion and hydrogen embrittlement.

Gray, H. R.↗
DOE OSTI2025

Development of Surveillance Test Articles for Materials Degradation Management in MSR Environments

Materials in molten salt reactors (MSR) undergo accelerated degradation from corrosion, irradiation, and cyclic loads at elevated temperatures. Establishing a materials surveillance program to enable the assessment of material deterioration is critical to assure structural integrity of MSRs components. This presentation summarizes recent efforts towards the development of surveillance test articles for collecting various damages for monitoring materials degradation. Surveillance test articles with reduced dimensions were designed to capture creep-fatigue damage from cyclic loading at elevated temperatures. The strain evolution of test articles during thermal cycling was analyzed both numerically and experimentally. Out-of-reactor thermal cycling demonstrated successful capture of strain range for materials assessment. Moreover, test articles were subject to both mechanical loads and molten salt exposure, and the damage due to stress and corrosion was investigated. Additionally, damage inference models were developed to predict the remaining life of materials based on the accumulated damage in surveillance test articles.

36 - MATERIALS SCIENCE↗
DOE OSTI2024

Effect of Engineered Cracks in Catalyst Layers on PEMFC Catalyst Layer Durability

Proton exchange membrane fuel cells (PEMFCs) are expected to play a pivotal role in decarbonizing the transportation sector, and particularly heavy-duty vehicles (HDVs). However, improvements in durability are needed for PEMFCs to compete with state-of-the-art power sources for HDVs. Here, we examine how catalyst layer (CL) cracks that are engineered affect the CL durability by using patterned silicon templates to control the CL crack density at the micrometer scale. Electrochemical analyses show that the initial PEMFC performance is relatively unaffected by crack density, but the performance after durability testing was strongly affected. Specifically, CLs with high crack density showed higher performance relative to CLs without cracks after application of a carbon corrosion accelerated stress test. Electrochemical analyses coupled with X-ray computed tomography and scanning transmission electron microscopy with energy dispersive X-ray spectroscopy showed that the cracks provide shorter oxygen diffusion pathways to reaction sites, leading to decreased oxygen transport resistance. Additionally, we observed that the catalyst durability is unaffected by cracks. Our results provide a mechanistic explanation of the role of cracks in CL durability.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2018

Variation of the Corrosion Behavior Prior to Crack Initiation of E690 Steel Fatigued in Simulated Seawater with Various Cyclic Stress Levels

The electrochemical process and corrosion behavior prior to crack initiation of E690 steel fatigued in simulated seawater are investigated under different cyclic stress levels by methods of electrochemical impedance spectroscopy, cyclic current response, electrochemical noise and morphology observation. The results show that the electrochemical behavior changes with the peak stress of cyclic loading and the corrosion behavior transforms as the peak stress increases over the proof strength. The charge transfer resistance (R{sub ct}) decreases with elevated peak stress. Continuous decreasing R{sub ct} with cycling number under peak stress of 1.02 σ{sub p0.2} indicates that continuous fresh metal induced by work-softening is exposed to the seawater, which is also proved by cyclic current response induced by cyclic stress and contact angle test. As a result, the corrosion type of E690 steel changes from general corrosion to pitting corrosion with the peak stress close to and above the proof strength. That induces the corrosion fatigue crack initiation mechanism to change correspondingly.

36 MATERIALS SCIENCE↗
DOE OSTI2012

SRNL phase ii shelf life studies - series 1 room temperature and high relative humidity

The Savannah River National Laboratory (SRNL) Phase II, Series 1 shelf-life corrosion testing for the Department of Energy Standard 3013 container is presented and discussed in terms of the localized corrosion behavior of Type 304 stainless steel in contact with moist plutonium oxide and chloride salt mixtures and the potential impact to the 3013 inner container. This testing was designed to address the influence of temperature, salt composition, initial salt moisture, residual stress and type of oxide/salt contact on the relative humidity inside a 3013 container and the initiation and propagation of localized corrosion, especially stress corrosion cracking.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Under the microscope: Reduced activation ferritic martensitic steel Eurofer-97 following ion-Irradiation and high-temperature high-pressure water exposure

Here, this study is designed to characterise the microstructural behaviour of Eurofer-97 steel under ion irradiation and subsequent exposure to high-temperature high-pressure (HTHP) water. Eurofer-97, a ferritic-martensitic steel, has been developed to withstand the conditions of fusion reactors in the locations in contact with coolant with an elevated level of neutron flux, such as the breeder-wall blanket. The material has been studied after self-ion irradiation (using Fe ions) simulating the microstructural effects of neutron irradiation limited to the subsurface layer. The corrosion properties of the Eurofer-97 steel were studied by exposure to HTHP water up to 331 °C. Advanced microstructural characterisation using scanning, transmission electron and focused ion beam microscopy was performed on the as-received microstructure and after ion irradiation. This was then characterised after exposure for 240 h in high-temperature water. Eurofer-97 had a dense, columnar Cr-rich inner oxide, followed by a Fe-rich outer oxide layer. In the irradiated condition the grain structure and oxide itself was less ordered. No appreciable difference in oxide thickness was identified between the irradiated and unirradiated specimens after this short exposure time.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Effect of water on critical and subcritical fracture properties of Woodford shale

Subcritical fracture behavior of shales under aqueous conditions is poorly characterized despite increased relevance to oil and gas resource development and seal integrity in waste disposal and subsurface carbon sequestration. In this paper, we measured subcritical fracture properties of Woodford shale in ambient air, dry CO 2 gas, and deionized water by using the double-torsion method. Compared to tests in ambient air, the presence of water reduces fracture toughness by 50%, subcritical index by 77%, and shear modulus by 27% and increases inelastic deformation. Comparison between test specimens coated with a hydrophobic agent and uncoated specimens demonstrates that the interaction of water with the bulk rock results in the reduction of fracture toughness and enhanced plastic effects, while water-rock interaction limited to the vicinity of the propagating fracture tip by a hydrophobic specimen coating lowers subcritical index and increases fracture velocity. Finally, the observed deviation of a rate-dependent subcritical index from the power law K-V relations for coated specimens tested in water is attributed to a time-dependent weakening process resulting from the interaction between water and clays in the vicinity of the fracture tip.

58 GEOSCIENCES↗
DOE OSTI2022

Machine learning in nuclear materials research

Nuclear materials are often demanded to function for extended time in extreme environments, including high radiation fluxes with associated transmutations, high temperature and temperature gradients, mechanical stresses, and corrosive coolants. They also have a wide range of microstructural and chemical makeups, resulting in multifaceted and often out-of-equilibrium interactions. Machine learning (ML) is increasingly being used to tackle these complex time-dependent interactions and aid researchers in developing models and making predictions, sometimes with better accuracy than traditional modeling that focuses on one or two parameters at a time. Conventional practices of acquiring new experimental data in nuclear materials research are often slow and expensive, limiting the opportunity for data-centric ML, but new methods are changing that paradigm. Here we review high-throughput computational and experimental data approaches, especially robotic experimentation and active learning that is based on Gaussian process and Bayesian optimization. We show ML examples in structural materials (e.g., reactor pressure vessel (RPV) alloys and radiation detecting scintillating materials) and highlight new techniques of high-throughput sample preparation and characterizations, and automated radiation/environmental exposures and real-time online diagnostics. Herein, this review suggests that ML models of material constitutive relations in plasticity, damage, and even electronic and optical responses to radiation are likely to become powerful tools as they develop. Finally, we speculate on how the recent trends of using natural language processing (NLP) to aid the collection and analysis of literature data, interpretable artificial intelligence (AI), and the use of streamlined scripting, database, workflow management, and cloud computing platforms that will soon make the utilization of ML techniques as commonplace as the spreadsheet curve-fitting practices of today.

36 MATERIALS SCIENCE↗
DOE OSTI2023

In-situ measurement and control of the tool-workpiece interface temperature during friction stir processing of 304/304L stainless steel

Friction stir processing of 304/304L stainless steel is an area of interest for repair of spent nuclear fuel canisters. To enable repair without canister deflection or fracture, it is necessary to define the limitations for the depth of the processing zone which will be critically dependent on the thermal history of the material. Here, this work provides the first direct measurement of the interface temperature for friction stir processing of steel. The measured temperature is in the range of 850–1050 °C which is 20–30 % higher than previously reported values. Notably, the peak temperature occurs at the pin tip which is attributed to a higher sticking fraction and subsequently higher strain rate near the tool axis of rotation. Furthermore, a clear correlation between the temperature, grain size, and hardness is established within a single weld nugget and across different processing conditions. In the future, this data may lead to improved process modelling through direct validation and allow for property optimization through direct microstructure control.

304 stainless steel↗