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 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Crack Growth Rate and Large Plate Demonstration of Chloride-induced Stress Corrosion Cracking in Spent Nuclear Fuel Storage Canisters

This report describes the body of work involving testing and analyses performed in FY18 and FY19 at the Savannah River National Laboratory and at Korea University to determine chloride-induced stress corrosion crack (CISCC) growth rates in weldments of austenitic stainless steel plate material. The primary work involved: several separate crack growth rate test campaigns with new testing setups using laboratory-scale corrosion specimens; analyses to estimate the stress intensity factors for flaws postulated to be present in the residual stress fields of axial and circumferential weldments in a prototypic dry storage canister; and establishment of a “large plate” (approximately 51×46 cm with thickness 16 mm and weight about 30 kg) CISCC test that uses a plate with a circumferential weldment harvested from the Sandia National Laboratories’ full-size mockup dry storage canister. The tests for crack growth rate were conducted on ASTM E1861 bolt-load compact tension specimens in a setup designed to allow initially dried salt deposits to deliquesce and infuse the brine to the crack front under conditions relevant to the canister storage environments (e.g., temperature and humidity). The crack growth rate test results matched previous experimental data in the open literature. These literature data had been compiled by SNL and EPRI to develop a temperature-dependent (K-insensitive) crack growth rate model to support flaw disposition in the ASME Boiler and Pressure Vessel Section XI Code Case N-860, “Examination Requirements and Acceptance Standards for Spent Nuclear Fuel Storage and Transportation Containment Systems.” The knowledge of the salt, temperature, humidity conditions for CISCC including the threshold stress intensity factor, $K_{ISCC}$, were used to setup the large plate test in which a set of partthrough- wall and through-wall electrical discharge machined (EDM) defects were positioned in the weld residual stress field of the welded plate material from the mockup canister. A salt solution was applied to this plate specimen with an air-brush spray, then dried, and then exposed to a constant 75% RH condition at an ambient room temperature of 22 °C. The salt loading and environmental condition is expected to be aggressive to cause CISCC. The visual results at the surface of the plate show corrosion product staining (brown rust appearance) but with no CISCC emanating from the machined notches after approximately 5 months exposure to date. The plate test will continue in FY20 and phased-array ultrasonic testing and other NDE techniques will be applied to identify and characterize CISCC that may emanate from the machined defects. The notches and any CISCC from them will be sectioned at the conclusion of the test to verify NDE characterization, and to develop effective through-wall crack growth rates from cracks that would likely be branched.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Evaluation of Nondestructive Examination Responses from Chloride-Induced Stress Corrosion Cracks: Fabrication of Base Metal Test Specimens

This report documents efforts to create actual chloride-induced stress corrosion cracks (CISCCs) in stainless-steel specimens for evaluating application of nondestructive examination (NDE) technologies for detection and sizing of CISCC in stainless-steel canisters in dry storage systems (DSSs) for spent nuclear fuel. In the United States, several DSSs for commercial spent nuclear fuel are approaching the end of their initial licensed or certified term. Many of these systems were originally licensed or certified for 20 years, after which the licensing or certification may be renewed for periods up to 40 years, according to Title 10 of the Code of Federal Regulations (10 CFR), Part 72, “Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater than Class C Waste.” Guidance for review of renewal applications is provided in NUREG-1927, Standard Review Plan for Renewal of Spent Fuel Dry Cask Storage System Licenses and Certificates of Compliance. A review of aging effects that could affect important-to-safety structures, systems, and components is required in renewal applications and applicants are required to address aging effects using either a time-limited aging analysis or an aging management program. The U.S. Nuclear Regulatory Commission (NRC) issued NUREG-2214, Managing Aging Processes In Storage (MAPS) Report, which provides guidance on the evaluation of known aging degradation mechanisms to determine if they could affect the ability of dry storage system components to fulfill their safety functions in the 20- to 60-year period of extended operation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Improved Measurement of Particulate Retention in Simulated Stress Corrosion Cracks

This report documents proposed improvements to an apparatus for measuring flow rates and aerosol retention in stress corrosion cracks (SCCs). The potential for SCCs in canister walls is a concern for dry cask storage systems for spent nuclear fuel. Some of the canisters in these systems are backfilled to significant pressures to promote heat rejection via internal convection. Pressure differentials covering the upper limit of commercially available dry cask storage systems are the focus of the current test assembly. Initial studies will be conducted using engineered microchannels with characteristic dimensions expected in SCCs that hypothetically could form in dry storage canister walls. In a previous study, an apparatus and procedures were developed and implemented to investigate aerosol retention in a simple microchannel with an SCC-like opening of 28.9 gm (0.00110 in.). The width was 12.7 mm (0.500 in.), and the length was 8.86 mm (0.349 in.). These initial results indicated 44% of the aerosols available for transmission were retained upstream of microchannel However, limitations in the aerosol instruments available at the time of the preliminary study introduced known biases into the measurements. While these biases were identified and quantified, their presence introduced unwanted degrees of freedom into the measurements and reduced accuracy. Because these aerosol particle sizers (APS) were limited to sampling at atmospheric pressure, a mass flow controller was used to supply the sample upstream of the crack to the APS. The average line loss across all particle sizes for this mass flow controller was 50%. The sample downstream of the crack was delivered via a mass flow meter and caused a line loss of 20%. Another source of bias was using separate (but identical) instruments to measure the aerosols upstream and downstream of the microchannel, which could register up to 40% different when measuring the same sample stream. The experience of conducting the preliminary study highlighted the need for improvements in the experimental approach that would eliminate these biases and benefit future studies. An aerosol analyzer has been identified and ordered that is ideally suited for this study and should substantially mitigate these biases. Moving forward in the near term, the same simple microchannel will be further investigated using the improved aerosol instrumentation. Additionally, an offset microchannel with a step in the flow path will be designed and fabricated for similar testing. Looking out further, the capability to produce and test laboratory generated SCCs will be developed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Mechanisms of Mitigating Chloride-Induced Stress Corrosion Cracking of Austenitic Steels by Laser Shock Peening

This study investigates the effect of laser shock peening (LSP) on the chloride-induced stress corrosion cracking (SCC) of 304 austenitic steels. LSP can induce a high compressive residual stress to a depth of 700 µm and plastic deformation structures of dislocations, deformation twins, and stacking faults. Constant-load SCC tests in MgCl2 solution suggested that LSP can retard the crack initiation and slow the crack growth. LSP-treated subsurface layers experience ductile fracture while the central regions exhibit intergranular SCC. The LSP-induced deformation structures may impede dislocation slips, while the LSP-induced compressive residual stress can lessen the stress intensity factor of crack tips and decrease the local stress for film rupture.

Materials Science↗
DOE OSTI2023

Irradiation-assisted Stress Corrosion Cracking of PWRirradiated Type 347 Stainless Steel (NSUF CINR 19-16567)

This document serves as the final report for NSUF Project 19-16567 (Reference 1) and its supporting contracts (References 2 and 3). It documents completion of testing of specimens fabricated as previously described in Reference 4. At a high level, the purpose of the testing is to generate data to support continued operation of light water reactors (LWR) and provide further understanding of irradiation-assisted stress corrosion crack (IASCC) initiation in baffle-former bolt materials. Baffle-former bolts that had been extracted from commercial pressurized water reactors (PWRs) were machined into miniature 4-point bend specimens and TEM samples at the Westinghouse Churchill Site hot cell facility. These samples were then shipped to the University of Michigan to undergo IASCC initiation testing in simulated PWR primary coolant containing either LiOH or KOH additions. Overall, 15 samples were tested, 8 in water with KOH additions and 7 in water with LiOH additions, until indications of cracking were observed. Crack initiation stress was similar in both water environments at 50-60% of the irradiated yield stress. Post-test examination of the cracking sites, including scanning electron microscopy and transmission electron microscopy, demonstrated that the crack initiation mechanism and crack initiation rates were similar between the two different water chemistries. A grain boundary oxidation and oxide fracture mechanism was observed to be responsible for crack initiation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Ultrasonic inspection of austenitic stainless steel welds with artificially produced stress corrosion cracks

Austenitic stainless steel welds and nickel alloy welds, which are widely used in nuclear power plants, present major challenges for ultrasonic inspection due to the grain structure in the weld. Large grains in combination with the elastic anisotropy of the material lead to increased scattering and affect sound wave propagation in the weld. This results in a reduced signal-to-noise ratio, and complicates the interpretation of signals and the localization of defects. Mechanized ultrasonic inspection was applied to study austenitic stainless steel test blocks with different types of flaws, including inter-granular stress corrosion cracks (IGSCC). The results show that cracks located in the heat affected zone of the weld are easily detected when inspection from both sides of the weld is possible. In cases of limited accessibility, when ultrasonic inspection can be carried out only from one side of a weld, it may be difficult to distinguish between signals from scattering in the weld and signals from cracks.

42 ENGINEERING↗
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 OSTI2020

Large Plate Experiment of Chloride-induced Stress Corrosion Cracking in Spent Nuclear Fuel Storage Canisters

This report provides a description of the testing (in progress at the Savannah River National Laboratory) to evaluate the corrosion behavior of a large section of welded plate material harvested from the Sandia National Laboratories’ full-size mockup of a spent nuclear fuel (SNF) dry storage canister. The details of the establishment of this plate test, the interim corrosion results (from visual examination), and the general plan for examination of the plate in CY21 to evaluate crack growth behavior (if any) are provided. This work is intended to inform aging management programs for SNF canisters under conditions of extended dry storage. A “large plate” (approximately 51 x 46 cm with thickness 16 mm and weight about 30 kg) and containing a circumferential weldment from the mockup is the test specimen. The knowledge of the salt, temperature, humidity conditions for chloride-induced stress corrosion cracking (CISCC) including the threshold stress intensity factor, K iscc , were used to setup the large plate test in which a set of part-through-wall and through-wall electrical discharge machined (EDM) defects were positioned in the weld residual stress field of the welded plate material from the mockup canister. A marine salt solution was applied to this plate specimen with an air-brush spray, then dried, and then exposed to a sustained 71-73% RH condition at an ambient room temperature of 22 °C. The salt loading and environmental condition is expected to be aggressive to cause CISCC

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Investigation of shot peening residual stress on alloy 718 leaf springs for primary water stress corrosion cracking analysis

This paper investigates the residual stress distributions from shot peening of Nickel Alloy 718 fuel assembly hold down spring material of PWR. Precision X-Ray Diffraction (XRD) testing has been performed to quantify residual stress versus depth on shot peened Nickel Alloy 718 samples that have been plastically deformed to different levels of stress. The results from these XRD tests have been used to benchmark a finite element model (FEM) that includes the effects of the residual compressive stress layer from shot peening. The benchmarked method of applying shot peening residual stresses can be applied to FEMs of Nickel Alloy 718 fuel assembly hold down springs to investigate the stress distributions in the springs at different loading conditions. XRD test results with a comparison to the benchmarked FEM are presented. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2024

Assessing Uncertainty in Modeling Stress Corrosion Cracking

This report summarizes the collaboration between Sandia National Laboratories (SNL) and the Nuclear Regulatory Commission (NRC) to improve the state of knowledge on chloride induced stress corrosion cracking (CISCC). The foundation of this work relied on using SNL’s CISCC computer code to assess the current state of knowledge for probabilistically modeling CISCC on stainless steel canisters. This work is presented as three tasks. The first task is exploring and independently comparing crack growth rate (CGR) models typically used in CISCC modeling by the research community. The second task is implementing two of the more conservative CGR models from the first task into SNL’s full CISCC code to understand the impact of the different CGR models on a full probabilistic analysis while studying uncertainty from three key input parameters. The combined work of the first two tasks showed that properly measuring salt deposition rates is impactful to reducing uncertainty when modeling CISCC. The work in Task 2 also showed how probabilistic CGR models can be more appropriate at capturing aleatory uncertainty when modeling SCC. Lastly, appropriate and realistic input parameters relevant for CISCC modeling were documented in the last task as a product of the simulations considered in the first two tasks.

36 MATERIALS SCIENCE↗
DOE OSTI2017

FY17 Status Report: Research on Stress Corrosion Cracking of SNF Interim Storage Canisters.

This progress report describes work done in FY17 at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. Work in FY17 refined our understanding of the chemical and physical environment on canister surfaces, and evaluated the relationship between chemical and physical environment and the form and extent of corrosion that occurs. The SNL corrosion work focused predominantly on pitting corrosion, a necessary precursor for SCC, and process of pit-to-crack transition; it has been carried out in collaboration with university partners. SNL is collaborating with several university partners to investigate SCC crack growth experimentally, providing guidance for design and interpretation of experiments.

36 MATERIALS SCIENCE↗
DOE OSTI2018

FY18 Status Report: SNL Research into Stress Corrosion Cracking of SNF Interim Storage Canisters

This progress report describes work done in FY18 at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). The work focuses on stress corrosion cracking (SCC), the only mechanism by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. Work in FY18 continued several studies initiated in FY17 that are aimed at refining the understanding of the chemical and physical environment on canister surfaces, and evaluating the relationship between chemical and physical environment and the form and extent of corrosion that occurs. The SNL canister environment work focused on evaluating the stability of sea-salt deliquescent brines on the heated canister surface; an additional opportunity to analyze dusts sampled from an inservice spent nuclear fuel storage canister also arose. The SNL corrosion work focused predominantly on pitting corrosion, a necessary precursor for SCC, and process of pit-to-crack transition. SNL is collaborating with several university partners to investigate SCC crack growth experimentally, providing guidance for design and interpretation of experiments. The scope of these efforts targets near-marine Independent Spent Fuel Storage Installation environments which are generally considered to be most aggressive for pitting and SCC. Work to define the chemical and physical environment that could develop on storage canister surfaces in near-marine environments included experiments to evaluate the thermal stability of magnesium chloride brines, representative of the first brines to form when sea-salts deliquesce, with the specific goal of understanding and interpreting results of sea-salt and magnesium chloride corrosion experiments carried out under accelerated conditions. The experiments showed that magnesium chloride brines, and by extension, low RH sea-salt deliquescent brines, are not stable at elevated temperatures, losing chloride via degassing of HC1 and conversion to Mg-hydroxychlorides and carbonates. The experiments were carried out on an inert substrate to eliminate the effects of corrosion reactions, simulating brine stabilities in the absence of, or prior to, corrosion. Moreover, analysis of salts recovered from actively corroding metal samples shows that corrosion also supports or drives conversion of magnesium chloride or sea-salt brines to less deliquescent salts. This process has significant implications on corrosion, as the secondary phases are less deliquescent than magnesium chloride; the conversion reaction results in decreases in brine volume, and potentially results in brine dry-out. The deliquescence properties of these reaction products will be a topic of active research in FY19.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

SNL Research into Stress Corrosion Cracking of SNF Dry Storage Canisters (FY19 Status Report)

This progress report describes work done in FY19 at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. Work in FY19 refined our understanding of the chemical and physical environment on canister surfaces and evaluated the relationship between chemical and physical environment and the form and extent of corrosion that occurs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Chemistry Envelope for Pitting and Stress Corrosion Cracking Mitigation

The present waste chemistry controls for the Hanford double-shell tanks (DSTs) were established in the 1980s in response to tank failures caused by stress corrosion cracking (SCC) at the DOE Savannah River Site (SRS). Portions of the SRS chemistry limits were coupled with corrosion testing results from Pacific Northwest Laboratory (PNL) and adopted for corrosion control of the DST waste at the Hanford Site. Since that time, the temperatures of many DST wastes have decreased to less than 50 °C as the radioactive material has decayed, the testing program in Argentina demonstrated that chloride ion in Hanford Site waste could cause pitting, and other work showed that dilute solutions could cause pitting at the liquid air interface (LAI). Moreover, future transfers from the single shell tanks and the waste treatment plant and process changes may shift the DST waste chemistries to a broader range of compositions with higher aggressive anion concentrations than the wastes that are presently stored. The decreasing temperatures led to the formulation of new chemistry controls for the minimization of SCC risks for operations below 50 °C in 2010, and these controls were partially adopted. The Tank Integrity Expert Panel (TIEP) Corrosion Subgroup (CSG) recommended that additional testing be performed to assess the potential for pitting and LAI corrosion. Since 2015, DNV-GL, SRNL, and the Hanford Site laboratory have investigated pitting and LAI corrosion at current and anticipated DST chemistries and Savannah River National Laboratory (SRNL) undertook a statistically based investigation of the role of nitrate and halide ion induced pitting corrosion. The objective was to develop a comprehensive waste chemistry envelope for the simultaneous minimization of the pitting and SCC risks caused by halide and nitrate ions. This report summarizes the technical basis for the new waste chemistry envelope to minimize the potential for both pitting and cracking corrosion induced by nitrate and halide ions, and the strategies used for the validation of the new controls.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Mitigate Stress Corrosion Cracking (SCC) in High-Strength Al castings (CRADA 508)

This project will develop approaches to overcome the barriers to simultaneous high-strength and high resistance to stress corrosion cracking (SCC) in Al casting alloys. Thus, success in this project will enable fabrication of light-weight components that are otherwise made of either heavier cast iron, or lower-strength Al alloys which need to be oversized (and hence, are heavier) to compensate for lower strength. This project will leverage the availability of advanced analytical instruments and scientific expertise available only through the LightMAT Consortium. If successful, this project has the potential to enable light-weighting of multiple components in light-duty cars and trucks across all makes and models.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Assessment of the Mechanical Stress Improvement Process for Mitigating Primary Water Stress Corrosion Cracking in Nickel Alloy Butt Welds in Piping Systems Approved for Leak-Before-Break

In recent years, operating experience has shown that Alloy 82/182/600 materials used in reactor coolant system (RCS) pressure boundaries of pressurized water reactors (PWRs) are susceptible to primary water stress corrosion cracking (PWSCC). Cracking can initiate at the inside surface of these materials, in part, because of tensile residual stresses introduced by welding. These materials are present in piping systems that were approved by the U.S. Nuclear Regulatory Commission (NRC) for leak before-break (LBB) before PWSCC was found in RCS dissimilar metal butt welds. The identification of PWSCC led to concerns regarding the potential effect of this degradation on existing LBB analyses. In response to these concerns, the NRC Office of Nuclear Regulatory Research initiated a program entitled, “PWSCC in Leak-Before-Break Systems,” and a follow-on program entitled, “Degradation in Extremely Low Probability of Rupture (xLPR) Systems.” Under these programs, Pacific Northwest National Laboratory (PNNL) is assessing the various strategies being used by industry to manage potential or existing PWSCC in susceptible welds in piping systems approved for LBB. The commercial nuclear power industry has implemented strategies to manage potential or existing PWSCC at Alloy 82/182 dissimilar metal welds (DMW) in PWRs. One general strategy consists of management by a combination of mitigation plus inspection, where several mitigation techniques have been used by industry. The other general strategy is to manage potential PWSCC by inspection alone. At the request of the NRC, the American Society of Mechanical Engineers (ASME) took actions to develop improvements to the existing Code requirements to address the potential for PWSCC in Class 1 PWR piping butt welds fabricated with Alloy 82/182 weld materials. The ASME developed Code Case N-770, “Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities, Section XI, Division 1.” It was realized that the Code Case needed to be revised to address specific examinations such as welds with a design life of less than 10 years and post-weld preservice surface examinations. In general, the revised Code Case (N-770-1) has requirements for inspection of unmitigated as well as mitigated Alloy 82/182 RCS butt welds. As such, specific inspection requirements for welds mitigated by the Mechanical Stress Improvement Process (MSIP®), the subject of this technical letter report (TLR), are contained in the Code Case. The NRC incorporated ASME Code Case N-770-1 by reference into §50.55a (76 FR 36232, p. 36278) in June 2011.

MSIP↗
DOE OSTI2015

A Qualitative Comparison of the C-Ring Test and the Jones Test as Standard Practice Test Methods for Studying Stress Corrosion Cracking in Ferritic Steels

Creep-strength-enhanced-ferritic (CSEF) steels have been widely implemented as water wall alloy materials in the coal-fired power industry for many years. The stress corrosion cracking (SCC) behavior of this class of materials is currently of significant interest to the industry due to recent failures. To better understand the test methods used to characterize SCC behavior in the laboratory, three representative CSEF alloys (T23, T24, and T92) were subjected to two SCC test protocols: the Jones Test set forth in DIN 50915, and the C-ring SCC test set forth in ASTM G38-01. Samples were tested in either the as-received (normalized + tempered) condition or in the normalized condition (quenched from 1065 C). Samples were exposed to aerated water in one test case and de-aerated water in a second test case for a period of 7 days at 200 C. It was found that for both test protocols, the normalized condition with aerated water led to severe cracking for all three alloys, whereas no evidence of cracking was found for the other conditions.

Thomson, Jeffery K↗