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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.

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At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Stress corrosion cracking of several solution heat-treated Mg–X alloys

Here we studied the stress corrosion cracking (SCC) behaviour of solution heat-treated as-cast Mg0.1Zr, Mg1Mn, Mg0.1Sr, Mg0.3Si, Mg5Sn, Mg5Zn and Mg0.3Ca in distilled water (DW) using the linearly increasing stress test (LIST). SCC susceptibility was related to the stress rate for all the Mg–X alloys except for Mg0.1Sr. At ~7.0 × 10 ₋4 MPa s ₋1 , Mg0.3Si and Mg5Sn were immune to SCC; Mg0.1Zr, Mg1Mn and Mg0.1Sr suffered medium SCC; Mg5Zn and Mg0.3Ca suffered relatively more serious SCC. At ~7.0 × 10 ₋5 MPa s ₋1 , Mg5Zn and Mg0.3Ca suffered the most serious transgranular SCC, whilst the other Mg–X alloys suffered medium SCC.

36 MATERIALS SCIENCE↗
NASA NTRS1975

Stress corrosion cracking of Ti-8Al-1 Mo-1V in molten salts

The stress corrosion cracking (SCC) behavior of Ti-8Al-1 Mo-1V has been studied in several molten salt environments. Extensive data are reported for the alloy in highly pure LiCl-KCl. The influence of the metallurgical heat treatment and texture, and the mechanical microstructure show similarities with aqueous solutions at lower temperature. The fracture path and cracking modes are also similar to that found in other environments. The influence of H2O and H(-) in molten LiCl-KCl lead to the conclusion that hydrogen does not play a major role in crack extension in this environment.

Smyrl, W. H.↗
NASA NTRS1981

An improved stress corrosion test medium for aluminum alloys

A laboratory test method that is only mildly corrosive to aluminum and discriminating for use in classifying the stress corrosion cracking resistance of aluminum alloys is presented along with the method used in evaluating the media selected for testing. The proposed medium is easier to prepare and less expensive than substitute ocean water.

Humphries, T. S.↗
DOE OSTI2015

Understanding the Risk of Chloride Induced Stress Corrosion Cracking of Interim Storage Containers for the Dry Storage of Spent Nuclear Fuel: Evolution of Brine Chemistry on the Container Surface.

Although the susceptibility of austenitic stainless steels to chloride-induced stress corrosion cracking is well known, uncertainties exist in terms of the environmental conditions that exist on the surface of the storage containers. While a diversity of salts is present in atmospheric aerosols, many of these are not stable when placed onto a heated surface. Given that the surface temperature of any container storing spent nuclear fuel will be well above ambient, it is likely that salts deposited on its surface may decompose or degas. To characterize this effect, relevant single and multi-salt mixtures are being evaluated as a function of temperature and relative humidity to establish the rates of degassing, as well as the likely final salt and brine chemistries that will remain on the canister surface.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Stress Corrosion Cracking of Ferritic Materials for Fossil Power Generation Applications

Creep strength enhanced ferritic (CSEF) steels Grades 23, 24, 91, and 92 have been widely implemented in the fossil fired industry for over two decades. The stress corrosion cracking (SCC) behavior of these materials with respect to mainstay Cr-Mo steels (such as Grades 11, 12 and 22) has not been properly assessed, particularly in consideration of recent reported issues of SCC in CSEF steels. This report details the results of Jones test exposures of a wide range of materials (Grades 11, 22, 23, 24, and 92), material conditions (as-received, improper heat treatments, normalized, weldments) and environments (salt fog; tube cleaning environments including decreasing, scale removal, and passivation; and high temperature water) to compare the susceptibility to cracking of these steels. In the as-received (normalized and tempered) condition, none of these materials are susceptible to SCC in the environments examined. However, in the hardened condition, certain combinations of environment and alloy reveal substantial SCC susceptibility.

ferritic↗
DOE OSTI2018

Historical Review of Stress Corrosion Cracking in Concentrated Uranium-Molybdenum Alloys

This report presents a historical overview of stress corrosion cracking (SCC) in uranium-molybdenum alloys, focusing primarily on bare U-10 wt. pct. Mo (U-10Mo). The fuel system being pursued by the United States High Performance Research Reactor (USHPRR) project incorporates a zirconium interlayer applied early in the process which likely serves as a protective coating. This alloy has been chosen as the high density uranium monolithic fuel for converting research reactors from high enriched uranium (HEU) to low enriched uranium (LEU) fuel for its combination of fuel density and gamma phase stability.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Intergranular Stress Corrosion Crack Initiation and Temperature Dependence of Alloy 600 in Pressurized Water Reactor Primary Water

Stress corrosion crack (SCC) initiation of a solution annealed, cold-worked (CW) UNS N06600 (Alloy 600) material was investigated in 360 and 325ºC simulated PWR primary water using constant load tensile instrumented for in-situ detection by the direct current potential drop technique. This CW material with high boron and low carbon bulk concentrations was found to be highly susceptible to SCC. Consistent SCC initiation times were obtained at both temperatures and the thermal activation energy for SCC initiation is estimated at ~116 kJ/mol.

Zhai, Ziqing↗
DOE OSTI2022

Crystallographic effects on transgranular chloride-induced stress corrosion crack propagation of arc welded austenitic stainless steel

Abstract The effect of crystallography on transgranular chloride-induced stress corrosion cracking (TGCISCC) of arc welded 304L austenitic stainless steel is studied on >300 grains along crack paths. Schmid and Taylor factor mismatches across grain boundaries (GBs) reveal that cracks propagate either from a hard to soft grain, which can be explained merely by mechanical arguments, or soft to hard grain. In the latter case, finite element analysis reveals that TGCISCC will arrest at GBs without sufficient mechanical stress, favorable crystallographic orientations, or crack tip corrosion. GB type does not play a significant role in determining TGCISCC cracking behavior nor susceptibility. TGCISCC crack behaviors at GBs are discussed in the context of the competition between mechanical, crystallographic, and corrosion factors.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Preliminary Modeling of Chloride Deposition on Spent Nuclear Fuel Canisters in Dry Storage Relevant to Stress Corrosion Cracking

This study presents a method that can be used to gain information relevant to determining the corrosion risk for spent nuclear fuel (SNF) canisters during extended dry storage. Currently, it is known that stainless steel canisters are susceptible to chloride-induced stress corrosion cracking (CISCC). However, the rate of CISCC degradation and the likelihood that it could lead to a through-wall crack is unknown. This study uses well-developed computational fluid dynamics and particle-tracking tools and applies them to SNF storage to determine the rate of deposition on canisters. The deposition rate is determined for a vertical canister system and a horizontal canister system, at various decay heat rates with a uniform particle size distribution, ranging from 0.25 to 25 µm, used as an input. In all cases, most of the dust entering the overpack passed through without depositing. Most of what was retained in the overpack was deposited on overpack surfaces (e.g., inlet and outlet vents); only a small fraction was deposited on the canister itself. These results are provided for generalized canister systems with a generalized input; as such, this paper is intended to demonstrate the technique. This study is a part of an ongoing effort funded by the U.S. Department of Energy, Nuclear Energy Office of Spent Fuel Waste Science and Technology, which is tasked with doing research relevant to developing a sound technical basis for ensuring the safe extended storage and subsequent transport of SNF. This work is being presented to demonstrate a potentially useful technique for SNF canister vendors, utilities, regulators, and stakeholders to utilize and further develop for their own designs and site-specific studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

US NRC-Sponsored Research on Stress Corrosion Cracking Susceptibility of Dry Storage Canister Materials in Marine Environments - 13344

At a number of locations in the U.S., spent nuclear fuel (SNF) is maintained at independent spent fuel storage installations (ISFSIs). These ISFSIs, which include operating and decommissioned reactor sites, Department of Energy facilities in Idaho, and others, are licensed by the U.S. Nuclear Regulatory Commission (NRC) under Title 10 of the Code of Federal Regulations, Part 72. The SNF is stored in dry cask storage systems, which most commonly consist of a welded austenitic stainless steel canister within a larger concrete vault or overpack vented to the external atmosphere to allow airflow for cooling. Some ISFSIs are located in marine environments where there may be high concentrations of airborne chloride salts. If salts were to deposit on the canisters via the external vents, a chloride-rich brine could form by deliquescence. Austenitic stainless steels are susceptible to chloride-induced stress corrosion cracking (SCC), particularly in the presence of residual tensile stresses from welding or other fabrication processes. SCC could allow helium to leak out of a canister if the wall is breached or otherwise compromise its structural integrity. There is currently limited understanding of the conditions that will affect the SCC susceptibility of austenitic stainless steel exposed to marine salts. NRC previously conducted a scoping study of this phenomenon, reported in NUREG/CR-7030 in 2010. Given apparent conservatisms and limitations in this study, NRC has sponsored a follow-on research program to more systematically investigate various factors that may affect SCC including temperature, humidity, salt concentration, and stress level. The activities within this research program include: (1) measurement of relative humidity (RH) for deliquescence of sea salt, (2) SCC testing within the range of natural absolute humidity, (3) SCC testing at elevated temperatures, (4) SCC testing at high humidity conditions, and (5) SCC testing with various applied stresses. Results to date indicate that the deliquescence RH for sea salt is close to that of MgCl{sub 2} pure salt. SCC is observed between 35 and 80 deg. C when the ambient (RH) is close to or higher than this level, even for a low surface salt concentration. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗