SEARCH · Search NASA
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.
Report details comprehensive study of nondestructive methods for detecting and/or evaluating up to six different types of corrosion: general, galvanic filiform, pitting, intergranular, and stress. Studied methods include: visual/optical inspection, radiography, penetrant inspection, analysis by polarized light, electrical conductivity measurements, pH analysis, and ultrasonic inspection.
A highly irradiated stainless steel 304 material irradiated to 27 dpa in the Experimental Breeder Reactor-II was used to investigate the relation between void swelling, grain boundary cohesion and intergranular crack growth rate under solicitation by irradiation assisted stress corrosion cracking. The material selected offered two different swelling amounts (about 2% and 3.7%) in the same material at high dose. The material had been irradiated at the same time, same condition, only the temperature gradient led to different swelling. This material, however, proved to be very resistant to irradiation assisted stress corrosion cracking under pressurized water reactor conditions. The crack growth rates measured were very low with the lowest being obtained with the material exhibiting the most amount of swelling. Micromechanical testing performed with the material exhibiting the higher crack propagation rate demonstrated that the irradiation damages (including void swelling) did not decrease grain boundary cohesion enough to make the grain boundaries the weak points under load solely. Crack path and crack tip analysis showed evidence of material modification ahead of the crack but it is not clear, at this point, if the changes observed were caused by the crack advance or if their presence in front of the crack was causing the low crack growth rate measured.
Stress corrosion crack initiation of two thermally-treated, cold-worked (CW) alloy 690 (UNS N06690) materials was investigated in 360oC simulated PWR primary water using constant load tensile (CLT) tests and blunt notch compact tension (BNCT) tests equipped with direct current potential drop (DCPD) for in-situ detection of cracking. SCC initiation was not detected by DCPD for either the 21% or 31%CW CLT specimens loaded at their yield stress after ~9,220 hours, however intergranular (IG) precursor damage and isolated surface cracks were observed on the specimens. The two 31%CW BNCT specimens loaded at moderate stress intensity after several cyclic loading ramps showed DCPD-indicated crack initiation after 10,400 hours of exposure at constant stress intensity, which was resulted from significant growth of IG cracks. The 21%CW BNCT specimens only exhibited isolated small IG surface cracks and showed no apparent DCPD change throughout the test. Post-test cross-section examinations revealed many grain boundary (GB) nano-cavities in the bulk of all the CLT and BNCT specimens particularly for the 31%CW materials. Cavities were also found along GBs extending to the surface suggesting an important role in crack nucleation. This paper provides an overview of the evolution of GB cavities and discusses their effects on crack initiation in CW alloy 690.
The increase in the penetration of tantalum by liquid lithium with increasing oxygen content of tantalum results from the ability of lithium to extract oxygen from tantalum oxides and/or from regions of high oxygen concentrations, accompanied by the creation of minute channels. At low oxygen concentration, oxides are present primarily at the grain boundaries and the lithium penetration is primarily intergranular. At high oxygen concentration, oxides and/or regions of high oxygen concentration are present both at the grain boundaries and within the grains; lithium penetration is both intergranular and transgranular. For a specific oxygen concentration, the severity of corrosion decreases with increasing temperature because (1) the oxygen solubility increases with temperature, decreasing the number of oxide particles and/or regions of high oxygen concentration, and (2) the diffusivity of oxygen in tantalum increases, allowing more rapid oxygen removal at the tantalum lithium interface.
This project focused on providing a fundamental physico-chemical understanding of the coupling mechanisms of corrosion- and radiation-induced degradation at material-salt interfaces in Ni-based alloys operating in emulated Molten Salt Reactor(MSR) environments through the use of a unique suite of aging experiments, in-situ nanoscale characterization experiments on these materials, and multi-physics computational models. The technical basis and capabilities described in this report bring us a step closer to accelerate the deployment of MSRs by closing knowledge gaps related to materials degradation in harsh environments.
This report outlines preliminary results on the survey and characterization of irradiated baffle plate, baffle former, and flux thimble tube specimens as part of the ongoing investigation into irradiation-induced embrittlement in austenitic stainless steels at ambient (room temperature) conditions. The specimens originated from commercial pressurized water reactor (PWR) components, covering displacement damage doses ranging from approximately 0.065 to 75 dpa. Initial scanning electron microscopy (SEM) surveys revealed that specimen surfaces exhibited fine machining marks and in-service-formed oxide layers on the side surfaces of the analyzed specimens. The oxide layers revealed specific features resembling localized "pitting-like" corrosion. An additional set of specimens representing in-service oxidation is being preserved for future microstructure analysis work. As believed, this will provide additional insights into long-term material degradation in PRWs. Specimen machining challenges emerged due to a complex failure of the electric discharge machining (EDM) system located in hot area. Despite partial restoration, issues persist with the EDM’s secondary power supply, necessitating the exploration of alternative EDM or computer numerical control (CNC) machining approaches to facilitate tensile specimen preparation. Currently, low speed saw cutting is in progress to prepare specimens for hydrogen measurements and general microstructure analysis. The near-term goals include completing tensile specimen machining for advanced mechanical testing and characterizing fracture mechanisms, and stress-corrosion cracking testing, ultimately aiming to identify and mitigate the ambient-condition intergranular cracking through targeted post-irradiation annealing strategies.
ORNL Molten Salt Reactor Experiment (MSRE) – 60’s- early 70’s Developed INOR-8 (aka Hastelloy N); MSRE operated over the period Dec 1965 to 1969. (not continuous); Operated for >13,000 hr at T from ~550 to ~650°C using U, Li-7, Be and Zr fluoride salt fuel; Post-Irradiation Material Evaluation (corrosion + neutron irradiation) of surveillance specimens; Noted ”beneficial effect of Nb” in alloy (on intergranular Te embrittlement) More recent (post-1980) Materials R&D for Gen IV …. Alloys 316L, 316H, 304, Alloy 800H, Alloy 617 (!!), Hastelloy N (mod), plus numerous experimental alloys developed
Scanning precession electron diffraction in the transmission electron microscope has been used to simultaneously map the phase, orientation, and grain morphology of oxides formed on Zircaloy-2 after three and six cycles in a boiling water reactor in unprecedented detail. For comparison, a region of a preoxidized autoclave-formed oxide was also proton-irradiated at the Dalton Cumbrian Facility. The proton irradiation was observed to cause additional stabilization of the tetragonal phase that was attributed to the stabilizing effect of irradiation- induced defects in the oxide. In the reactor-formed oxides, no extra stabilization of the tetragonal grains was observed under neutron irradiation, as indicated by the similar tetragonal phase fraction and transformation twin-boundary distributions between the nonirradiated and reactor-formed oxides. It is suggested that the damage rate is too low in the newly formed oxide to cause significant stabilization of the tetragonal phase. This technique also reveals that the oxide formed under reactor conditions has a more heterogeneous microstructure, and the growth of well-oriented columnar monoclinic grains is significantly reduced compared with a nonirradiated oxide. High-angle annular dark-field scanning transmission electron microscopy also revealed the development of extensive networks of intergranular porosity and eventually grain de-cohesion in the reactor-formed oxides. These results suggest that the tetragonal-monoclinic transformation is not responsible for the accelerated corrosion exhibited under reactor conditions. It is proposed that the usual out-of-reactor oxide growth and nucleation processes are significantly modified under reactor conditions, resulting in a more heterogeneous and randomly oriented oxide microstructure with reduced columnar grain growth. It is suggested that this disordered oxide microstructure allows for the formation of extensive intergranular porosity that could lead to accelerated in-reactor corrosion. (authors)
Selective oxidation is a critical component of the proposed internal oxidation mechanism of stress corrosion cracking (SCC) for Ni-based alloys in primary water reactor (PWR) operation. To isolate and study this process, the binary Ni-20Cr was exposed to pressurized water and a CO/CO2 gas mixture. In both environments, the oxygen potential was kept under the Ni/NiO line to stimulate penetrative intergranular attack. A combination of electron microscopy and atom probe tomography was used to study the nanometer-scale details of the passivation and penetrative intergranular oxidation processes at high-energy grain boundaries. Oxygen transport towards the terminating oxidation front is elucidated with dedicated usage of oxygen isotopic tracer exchange experiments to support classical theories of internal oxidation, revealing preferred oxygen transport paths were visualized with sub-nanometer resolution.
The effect of temper condition and corrosion on the fatigue behavior of alloy AA2219 has been investigated in different temper conditions (T87 and T851). Corrosion testing was performed by exposing the tensile specimens to 3.5% NaCl solution for different time periods, and the corrosion damage was quantified using a 3D profilometer. The exposure-tested specimens were subjected for fatigue testing at different stress levels, and the reduction in fatigue life was measured along with detailed fracture morphology variations. The results indicated that the alloy in both tempers suffers localized corrosion damage and the measured corrosion depth was 120 and 1200 µm, respectively, for T87 and T851 conditions. The loss in fatigue strength was found to be high for T851 (67%) when compared to that of T87 temper condition (58%) for a pre-corrosion time of 15 days. In both cases, fatigue crack initiation is associated with corrosion pits, which act as stress raisers. However, the crack propagation was predominantly transgranular for T87 and a mixed transgranular and intergranular fracture in the case of T851 temper condition. This was shown to be due to the heterogeneous microstructure due to the thermomechanical working and the delay in quench time imposed on the alloy forging in T851 temper condition. The findings in this paper present useful information for the selection of appropriate heat treatment condition to facilitate control of the corrosion behavior which is of great significance for their fatigue performance.
Performance degradation of ternary layered oxide cathodes largely originates from their loss of structural integrity in cyclic usage. Mechanical damage, such as intergranular fracture of the active particles, is not only a mechanical cleavage process but also interferes with electrochemical kinetics such as infiltration of liquid electrolyte, surface corrosion of the constituent primary particles, and may eventually isolate the primary grains from the electron conducting network. Here, in this work, we develop a computational framework that integrates electrochemistry of a LiNi x Mn y Co 1−x−y O 2 (NMC) composite cathode with mechanical damage of the active particles. To fully examine the intricate chemomechanical behavior of the electrode, we evaluate the effects of the anisotropic material properties, the influence of mechanical potential on Li transport, and the concurrent intergranular fracture and electrolyte penetration along the grain boundaries upon multiple cycles. Electrolyte infiltration benefits capacity retention but aggravates further mechanical damage by corrosion. Structural failure mostly occurs in the first charging due to the anisotropic mechanical strain between the primary grains, while the resulting damage remains stable in the later few cycles. The results are consistent with experimental observations and the integration of electrochemistry and mechanical failure enables a step further understanding of the complex mechanism of battery degradation.
Stress corrosion crack (SCC) initiation of three mill-annealed (MA) alloy 600 heats in simulated pressurized water reactor primary water has been investigated using constant load tests equipped with in-situ direct current potential drop (DCPD) measurement capabilities. SCC initiation times were greatly reduced by a small amount of cold work. Shallow intergranular (IG) attack and/or cracks were found on most high-energy grain boundaries intersecting the surface with only a small fraction evolving into larger cracks and IGSCC growth. Crack depth profiles were measured and related to DCPD-detected initiation response. Processes controlling the SCC initiation in MA alloy 600 are discussed. IN PRESS, CORRECTED PROOF, 05/02/2017 - mfl
Detailed destructive examinations were conducted to characterize the integrity and material properties of two aluminum alloy (7075-T6) horizontal stabilizer tie box forgings removed.from US. Air Force C5A and C5B transport aircraft. The C5B tie box forging was,found to contain no evidence of cracking. Thirteen cracks were found in the CSA,forging. All but one of the cracks observed in the C5A component were located along the top cap region (one crack was located in the bottom cap region). The cracks in the C5A component initiated at fastener holes and propagated along a highly tunneled intergranular crack path. The tunneled crack growth configuration is a likelv result of surface compressive stress produced during peening of the .forging suijace. The tie box forging ,fatigue crack growth, fracture and stress corrosion cracking (SCC) properties were characterized. Reported herein are the results of laboratory air ,fatigue crack growth tests and 95% relative humidity SCC tests conducted using specimens machined from the C5A ,forging. SCC test results revealed that the C5A ,forging material was susceptible to intergranular environmental assisted cracking: the C5A forging material exhibited a SCC crack-tip stress-intensity factor threshold of less than 6 MPadn. Fracture toughness tests revealed that the C5A forging material exhibited a fracture toughness that was 25% less than the C5B forging. The C5A forging exhibited rapid laboratory air fatigue crack growth rates having a threshold crack-tip stress-intensity factor range of less than 0.8 MPa sup m. Detailed fractographic examinations revealed that the ,fatigue crack intergranular growth crack path was similar to the cracking observed in the C5A tie box forging. Because both fatigue crack propagation and SCC exhibit similar intergranular crack path behavior, the damage mechanism resulting in multi-site cracking of tie box forgings cannot be determined unless local cyclic stresses can be quantified.
High-resolution microscopy of a high-purity Ni-5Cr alloy exposed to 360°C hydrogenated water reveals intergranular selective oxidation of Cr accompanied by local Cr depletion and diffusion-induced grain boundary migration (DIGM). The corrosion-product oxide consists of a porous, interconnected network of Cr2O3 platelets with no further O ingress into the metal ahead. Extensive grain boundary depletion of Cr (to <0.05at.%) is observed typically 20–100 nm wide as a result of DIGM and reaching depths of many micrometers beyond the oxidation front.
The solid to liquid phase transformation of LiOH has been proposed as an ideal candidate thermal energy storage media for a Rankine Cycle powered electrical generation unit envisioned in Space Station based solar dynamic systems. Due to the corrosive nature of molten hydroxides, long term containment of LiOH is of concern. Pure nickel is thought to be a suitably resistant material, and a program has been instituted to measure the effects of prolonged exposure of liquid and gaseous LiOH on the mechanical properties of pure nickel alloys. Results to date indicate that negligible weight and thickness changes occurred in Ni alloys exposed to LiOH for as long as 2500 hr at 775 K, and essentially no difference in 77-900 K tensile properties could be detected between LiOH exposed and vacuum annealed Ni specimens. Although there was little sign of outward damage, microstructural examination revealed that all hydroxide contaminated tensile test specimens had surface connected intergranular cracks along the gage lengths. Two other potential problems, which have strong implications with respect to a LiOH/Ni energy storage system, were also noted during the corrosion experiments. In particular stress corrosion cracking of weld joints in pressurized vessel and permeation of hydrogen through nickel were observed.
We report after exposure to oxygen-poor (10 -13 –10 -14 wt%) liquid lead-bismuth eutectic (LBE) at 500 °C for 500 h, LBE penetrates more than one order of magnitude deeper in an FCC Al 0.4 CoCrFeNi high-entropy alloy (HEA) decorated with a network of BCC (Ni, Al)-rich intergranular (IG) precipitates than in a single-phase, FCC Al 0.3 CoCrFeNi HEA without the IG precipitate network. This deterioration of corrosion resistance is attributed to the energetic nature of the BCC/FCC interphase boundaries (IBs) and resultant IB wetting. The LBE ingress film selectively leaches nickel located at those low-indexed crystalline planes, resulting in phase transformation from FCC to BCC structure.
To investigate the impact of low-temperature post-irradiation annealing (PIA) on the stress corrosion cracking of neutron-irradiated 304 stainless steel, constant extension rate tests were conducted in simulated boiling water reactor normal water chemistry (BWR-NWC) and hydrogenated water chemistry (HWC) environments. Ten tensile samples, comprising five as-irradiated and five PIA specimens, were prepared by electropolishing the gauge section of electric discharge machined (EDM) samples. Here, the annealing treatment reduced the yield strength from approximately 550 MPa to around 425 MPa and significantly restored the ductility and the strain hardening capability of the alloy. Consequently, the susceptibility of this material to irradiation-assisted stress corrosion cracking (IASCC) was effectively mitigated, which is more prominent in HWC, as evident from fractography, which indicated a decreased propensity for intergranular (IG) fracture. Furthermore, it was observed that the polished surface facilitated crack initiation more readily than the EDM surface, suggesting that the EDM process suppressed crack initiation to some extent.
We report long-term laboratory scale static alteration tests (up to 1500 d) were conducted on two low-activity waste glasses (LAWA44 and IDF18-A161) at S/V = 2000 m –1 . Tests were seeded with either analcime, clinoptilolite, zeolite P1, or zeolite P2 to induce Stage III glass alteration behavior. The effect of several parameters on glass Stage III behavior was studied with consideration of several variables, including temperature (22 °C – 90 °C) and pH (floating or initially set using KOH at values from 9.5 to 11.5). Temporal solution data and solids characterization at the end of the experiments are reported. Solution data demonstrated the occurrence of a sustained rate acceleration at temperatures down to 40 °C and the activation energy for these Stage III rates was determined for both glasses. Three major secondary phases were identified at the end of the experiments for both glasses: zeolite P1, zeolite P2, and/or analcime (along with minor phases: zeolites, clays, carbonates). The identification of the crystalline phases at the end of the experiment often differed from the zeolite seed phase. In addition, the use of the Avrami equation showed unimpeded zeolite growth (3-dimensional) in certain conditions, suggesting that in certain conditions the glass dissolution rate is controlled by zeolite formation kinetics whereas in other conditions the glass corrosion process impeded zeolite formation kinetics. The morphological evolution of the samples revealed the growth of the zeolites which completely fill the intergranular space in the powder bed. These results are discussed with regards to previous seeded and unseeded Stage III dissolution rate experiments performed on low- and high-activity nuclear waste glasses.