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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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At least 613 records · Page 34
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Sub-critical crack growth in silicate glasses: Role of network topology

The presence of water in the surrounding atmosphere can cause sub-critical crack growth (SCCG) in glasses, a phenomenon known as fatigue or stress corrosion. Here, to facilitate the compositional design of more fatigue-resistant glasses, we investigate the composition dependence of SCCG by studying fourteen silicate glasses. The fatigue curves (V-K{sub I}) have been obtained by indentation experiments through measurements of the crack length as a function of post-indentation fatigue duration. Interestingly, we find that the fatigue resistance parameter N is generally improved by increasing the alumina content and is thereby found to exhibit a fairly linear dependence on the measured Vickers hardness H{sub V} for a wide range of N and H{sub V} values. This finding highlights the important role of network topology in governing the SCCG in silicate glasses, since hardness has been shown to scale linearly with the number of atomic constraints. Our results therefore suggest that glasses showing under-constrained flexible networks, which feature floppy internal modes of deformation, are more readily attacked by water molecules, thus promoting stress corrosion and reducing the fatigue resistance.

36 MATERIALS SCIENCE↗
NASA NTRS1977

Evaluation and control of environmental corrosion for aluminum and steel alloys

Corrosion protection systems for aerospace application and the effects of surface treatments and methods of controlling stress corrosion are evaluated. Chromate pigmented systems were found to be most effective for aluminum alloys; zinc-rich coatings gave the greatest protection to steel alloys. Various steel and aluminum alloys are rated for stress corrosion resistance.

Franklin, D. B.↗
DOE OSTI2014

Thermal Recovery of Plastic Deformation in Dissimilar Metal Weld

Stainless steel has been widely used in challenging environments typical to nuclear power plant structures, due its excellent corrosion resistance. Nickel filler metals containing high chromium concentration, including Alloy 82/182, are used for joining stainless steel to carbon steel components to achieve similar high resistance to stress corrosion cracking. However, the joint usually experience weld metal stress corrosion cracking (SCC), which affects the safety and structural integrity of light water nuclear reactor systems. A primary driving force for SCC is the high tensile residual stress in these welds. Due to large dimension of pressure vessel and limitations in the field, non-destructive residual stress measurement is difficult. As a result, finite element modeling has been the de facto method to evaluate the weld residual stresses. Recent studies on this subject from researchers worldwide report different residual stress value in the weldments [5]. The discrepancy is due to the fact that most of investigations ignore or underestimate the thermal recovery in the heat-affect zone or reheated region in the weld. In this paper, the effect of heat treatment on thermal recovery and microhardness is investigated for materials used in dissimilar metal joint. It is found that high equivalent plastic strains are predominately accumulated in the buttering layer, the root pass, and the heat affected zone, which experience multiple welding thermal cycles. The final cap passes, experiencing only one or two welding thermal cycles, exhibit less plastic strain accumulation. Moreover, the experimental residual plastic strains are compared with those predicted using an existing weld thermo-mechanical model with two different strain hardening rules. The importance of considering the dynamic strain hardening recovery due to high temperature exposure in welding is discussed for the accurate simulation of weld residual stresses and plastic strains. In conclsuion, the experimental result reveals that the typical post-buttering heat treatment for residual stress relief may not be adequate to completely eliminate the residual plastic strains in the buttering layer.

36 MATERIALS SCIENCE↗
DOE OSTI2025

XFEM Development for Modeling Crack Growth in Prototypical Welded Components

Nuclear power plant components are subjected to harsh operating environments that can lead to multiple degradation mechanisms in which fracture can play a prominent role. Predicting crack growth is important for assessing the integrity of welded components. The extended finite element method (XFEM) is an important tool for modeling such crack growth, and XFEM capabilities have been developed within the MOOSE framework. This report documents work in the MOOSE XFEM module to model fractures in three-dimensional representations of components using a topologically two-dimensional mesh to define cutting planes. Crack growth algorithms have been implemented to evolve the cutting mesh based on equations for stress corrosion cracking. Additionally, several usability and robustness improvements have been developed to enable three-dimensional fracture simulations. The cutting algorithms were demonstrated on a three-dimensional model of a prototypical reactor component undergoing stress corrosion cracking driven by idealized weld residual stresses. This is an incremental step toward using this capability to model more complex components with residual stresses computed through welding process simulations.

42 - ENGINEERING↗
DOE OSTI2015

Meso-scale magnetic signatures for nuclear reactor steel irradiation embrittlement monitoring

Verifying the structural integrity of passive components in light water and advanced reactors will be necessary to ensure safe, long-term operations of the existing U.S. nuclear fleet. This objective can be achieved through nondestructive condition monitoring techniques, which can be integrated with plant operations to quantify the “state of health” of structural materials in real-time. While nondestructive methods for monitoring many classes of degradation (such as fatigue or stress corrosion cracking) are relatively advanced, this is not the case for degradation caused by irradiation. The development of nondestructive evaluation technologies for these types of degradation will require advanced materials characterization techniques and tools that enable comprehensive understanding of nuclear reactor material microstructural and behavioral changes under extreme operating environments. Irradiation-induced degradation of reactor steels causes changes in their microstructure that impacts their micro-magnetic properties. In this paper, we describe preliminary results of integrating advanced material characterization techniques with meso-scale computational models. In the future, this will help to provide an interpretive understanding of the state of degradation in structural materials. Microstructural data are presented from monocrystalline Fe and are correlated with variable-field magnetic force microscopy and micro-magnetic measurements. Ongoing research is focused on extending the measurements and models on thin films to gain insights into the structural state of irradiated materials and the resulting impact on magnetic properties. Preliminary conclusions from these correlations are presented, and next steps described.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Final Scientific/Technical Report, Multi-Sensor Inspection and Robotic Systems for Dry Storage Casks

Once removed from service, fuel rod assemblies are cooled in a pool and then placed in dry storage until they can be transported to a repository for disposal. The extended duration that used nuclear fuel spends in dry storage makes it prudent to be able to nondestructively inspect the dry storage casks. The degradation mechanisms of most concern are chloride-induced stress corrosion cracking of the stainless steel canister that confines the spent fuel and distributed cracking of the concrete in the overpack structure. Multi-sensor inspection and robotic systems were researched, developed, built, and tested. One system operates completely outside the cask to inspect the concrete. The other system, known as the PRINSE system, for Proactive Robotic Inspection of Nuclear Storage Enclosures, enters the cask through an outlet vent and navigates around the outer surface of the stainless steel canister. The PRINSE system delivers sensors for: (i) real-time characterization of the canister surface in order to assess whether chlorides are present, (ii) detecting the presence of cracks using shear-horizontal ultrasonic guided waves, (iii) measuring the temperature of the canister surface, and (iv) a gamma radiation counter. The four sensing modalities are applicable to any cask containing a stainless steel canister, however the robotic delivery system is designed specifically for Holtec’s HI-STORM family of casks. Horizontal axis casks have exposed concrete surfaces. A fully contactless air-coupled scanning ultrasonic test platform and a frequency-wavenumber domain data analysis method are used for these concrete structures. Analysis, numerical simulation and experimental results demonstrate that distributed cracks in concrete can be detected and cracked regions localized within a scanned region of concrete. Vertical axis casks (like the HI-STORM family) have steel-clad concrete overpacks. In a separate effort, a semi-coupled test platform and a guided wave attenuation-based data analysis method are used for disbond detection. Analysis, numerical simulation and experimental results show that the bonding conditions of steel-concrete interfaces can be quantitatively characterized using the guided wave signal analysis scheme. The PRINSE system is designed for vertical axis dry storage casks, specifically the HI-STORM family. The double pulse laser-induced breakdown spectroscopy (LIBS) system takes advantage of an optical fiber to deliver intense light to the canister surface that enables surface composition characterization in-situ in real time. Chlorine concentrations as low as 10 mg/m2 were detected on stainless steel. Additionally, a simpler laser setup was created to detect sodium as a surrogate for chlorine, and sodium was detectable when the chlorine concentration was as low as 0.5 mg/m2. The electromagnetic acoustic transducers (EMATs) enable detection of stress corrosion cracks. The EMATs generate shear-horizontal waves that propagate around the canister and are reflected by cracks enabling 100% of the canister weld lines to be inspected. The EMAT-based crack detection system was tested at the Pacific Northwest National Laboratory on a canister containing blind defects and was shown to detect the crack sizes it was designed to detect. The PRINSE system is a novel marsupial robot. The command center and winch outside the cask control the delivery arm and train of cars that traverse the outside of the canister from the cask ventilation system. The sensors are housed in the cargo bays of three cars that comprise the train, which is lowered to the bottom of the canister by the winch. Encoders on the winch enable the sensors to be delivered to a prescribed location, e.g., for re-inspection of a problem area. The PRINSE system does not mar the canister surface, can be extracted manually in the case of a power failure, and is designed for an elevated temperature and gamma radiation environment, although not all components of the prototype are made from the final materials. The PRINSE system was tested on a full canister that was placed inside a HI-STORM cask at the Holtec Manufacturing Division. It was also demonstrated on a full-height mockup at Penn State to a live audience.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Analysis of Dust and Corrosion Witness Samples Recovered from SNF Dry Storage Systems, Maine Yankee, 2023

This report documents the results of a long-term (5.79 year) exposure of 4-point bend corrosion test samples in the inlet and outlet vents of four spent nuclear fuel dry storage systems at the Maine Yankee Independent Spent Fuel Storage Installation. The goal of the test was to evaluate the corrosiveness of salt aerosols in a realistic near-marine environment, providing a data set for improved understanding of stress corrosion cracking of spent nuclear fuel dry storage canisters. Examination of the samples after extraction showed minor corrosion was present, mostly on rough-ground surfaces. However, dye penetrant testing showed that no SCC cracks were present. Dust collected on coupons co-located with the corrosion specimens was analyzed by scanning electron microscopy and leached to determine the soluble salts present. The dust was mostly organic material (pollen and stellate trichomes), with lesser detrital mineral grains. Salts present were a mix of sea-salts and continental salts, with chloride dominating the anions, but significant amounts of nitrate were also present. Both corrosion samples and dust samples showed evidence of wetting, indicating entry of water into the vents. The results of this field test suggest that the environment at Maine Yankee is not highly aggressive, although extrapolation from the periodically wetted vent samples to the hot, dry, canister surface may be difficult. No stress corrosion cracks were observed, but minor corrosion was present despite high nitrate concentrations in the salts. These observations may help address the ongoing question of the importance of nitrate in suppressing corrosion and SCC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1987

Accelerated crack growth, residual stress, and a cracked zinc coated pressure shell

During a partial inspection of a 42 year old, operating, pressurized wind tunnel at NASA-Ames Research Center, a surface connected defect 114 in. long having an indicated depth of a 0.7 in. was detected. The pressure shell, constructed of a medium carbon steel, contains approximately 10 miles of welds and is cooled by flowing water over its zinc coated external surface. Metallurgical and fractographic analysis showed that the actual detect was 1.7 in. deep, and originated from an area of lack of weld penetration. Crack growth studies were performed on the shell material in the laboratory under various loading rates, hold times, and R-ratios with a simulated shell environment. The combination of zinc, water with electrolyte, and steel formed an electrolytic cell which resulted in an increase in cyclic crack growth rate by as much as 500 times over that observed in air. It was concluded that slow crack growth occurred in the pressure shell by a combination of stress corrosion cracking due to the welding residual stress and corrosion fatigue due to the cyclic operating stress.

Dittman, Daniel L.↗
DOE OSTI2014

Liquid-Air Interface Corrosion Testing Simulating The Environment Of Hanford Double Shell Tanks

Coupon tests on A537 carbon steel materials were conducted to evaluate the Liquid-Air Interface (LAI) corrosion susceptibility in a series of solutions designed to simulate conditions in the radioactive waste tanks located at the Hanford Nuclear Facility. The new stress corrosion cracking requirements and the impact of ammonia on LAI corrosion were the primary focus. The minimum R value (i.e., molar ratio of nitrite to nitrate) of 0.15 specified by the new stress corrosion cracking requirements was found to be insufficient to prevent pitting corrosion at the LAI. The pH of the test solutions was 10, which was actually less than the required pH 11 defined by the new requirements. These tests examined the effect of the variation of the pH due to hydroxide depletion at the liquid air interface. The pits from the current testing ranged from 0.001 to 0.008 inch in solutions with nitrate concentrations of 0.4 M and 2.0 M. The pitting and general attack that occurred progressed over the four-months. No significant pitting was observed, however, for a solution with a nitrate concentration of 4.5 M. The pitting depths observed in these partial immersion tests in unevaporated condensates ranged from 0.001 to 0.005 inch after 4 months. The deeper pits were in simulants with low R values. Simulants with R values of approximately 0.6 to 0.8 appeared to significantly reduce the degree of attack. Although, the ammonia did not completely eliminate attack at the LAI, the amount of corrosion in an extremely corrosive solution was significantly reduced. Only light general attack (< 1 mil) occurred on the coupon in the vicinity of the LAI. The concentration of ammonia (i.e., 50 ppm or 500 ppm) did not have a strong effect.

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

Effects of Radiation-Induced Defects on Corrosion

The next generation of nuclear reactors will expose materials to conditions that, in some cases, are even more extreme than those in current fission reactors, inevitably leading to new materials science challenges. Radiation-induced damage and corrosion are two key phenomena that must be understood both independently and synergistically, but their interactions are often convoluted. In the light water reactor community, a tremendous amount of work has been done to illuminate irradiation-corrosion effects, and similar efforts are under way for heavy liquid metal and molten salt environments. While certain effects, such as radiolysis and irradiation-assisted stress corrosion cracking, are reasonably well established, the basic science of how irradiation-induced defects in the base material and the corrosion layer influence the corrosion process still presents many unanswered questions. In this review, we summarize the work that has been done to understand these coupled extremes, highlight the complex nature of this problem, and identify key knowledge gaps.

36 MATERIALS SCIENCE↗
NASA NTRS1983

Dynamic fatigue of a machinable glass-ceramic

To assess the stress-corrosion susceptibility of a machinable glass-ceramic, its dynamic fatigue behavior was investigated by measuring its strength as a function of stress rate. Fracture mechanics techniques were used to analyze the results for the purpose of making lifetime predictions for components of this material. This material was concluded to have only moderate resistance (N = 30) to stress corrosion in ambient conditions. The effects of specimen size on strength were assessed for the material used in this study; it was concluded that the Weibull edge-flaw scaling law adequately describes the observed strength-size relation.

Smyth, K. K.↗
NASA NTRS1986

Aqueous sodium chloride induced intergranular corrosion of Al-Li-Cu alloys

Two methods have been explored to assess the susceptibility of Al-Li-Cu alloys to intergranular corrosion in aqueous sodium chloride solution. They are: (1) constant extension rate testing with and without alternate-immersion preexposure and (2) metallographic examination after exposure to a NaCl-H2O2 corrosive solution per Mil-H-6088F. Intergranular corrosion was found to occur in both powder and ingot metallurgy alloys of similar composition, using both methods. Underaging rendered the alloys most susceptible. The results correlate to stress-corrosion data generated in conventional time-to-failure and crack growth-rate tests. Alternate-immersion preexposure may be a reliable means to assess stress corrosion susceptibility of Al-Li-Cu alloys.

Pizzo, P. P.↗
DOE OSTI2019

Corrosion Monitoring in an Irradiated/Used Fuel Dry Storage Container - 19193

A corrosion monitoring system was developed to monitor the corrosion behaviour of steel components embedded in the high density concrete of a Dry Storage Container (DSC). The DSC is a free-standing reinforced concrete container, with an inner carbon-steel liner and an outer carbon-steel shell, for the storage, on-site transfer and off-site transportation of irradiated Candu fuel. Normally, uncontaminated concrete is benign with regards to embedded steel components because high alkalinity in the concrete leads to formation of a passive film on the steel surface, which prevents excessive metal loss. However, corrosion of steel in concrete may be promoted by one or more of local concentration of chloride ions or local loss of alkalinity. For the DSC, it is recognized that the above mechanism(s) are unlikely, however, if any did exist, the corrosion mechanisms that could then occur would include corrosion of carbon steel components and stress corrosion cracking of 304L stainless steel components. In order to confirm that these unlikely events are not occurring, Ontario Power Generation has undertaken a project to investigate the corrosion behaviour of metallic components embedded in the DSC's high density concrete. A corrosion monitoring system that can be installed within the concrete space of a DSC was designed and installed to monitor changes in temperature, corrosion potential, corrosion rate, and electrochemical noise. The corrosion monitoring system was commissioned in July 2012, followed by the commencement of monitoring and measurements. The corrosion data, collected for the three years from year 2012 to 2015, were analyzed. It was concluded that the corrosion rates of both carbon steel and stainless steel within the DSC are very low and consistent with the intended minimum operational life of the DSC. The monitoring system as a whole has allowed a systematic and long term judgement regarding the conditions within the DSCs. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1975

Investigation of smooth specimen scc test procedures; variations in environment, specimen size, stressing frame, and stress state

The variables studied in the stress-corrosion cracking performance of high strength aluminum alloys were: (1) corrosiveness of the environment, (2) specimen size and stiffness of the stressing system, (3) interpretation of transgranular cracking, and (4) interaction of the state of stress and specimen orientation in a product with an anisotropic grain structure. It was shown that the probability of failure and time to fracture for a specimen loaded in direct tension are influenced by corrosion pattern, the stressing assembly stiffness, and the notch tensile strength of the alloy. Results demonstrate that the combination of a normal tension stress and a shear stress acting on the plane of maximum susceptibility in a product with a highly directional grain cause the greatest tendency for stress-corrosion cracking.

Lifka, B. W.↗
DOE OSTI2014

A FRAMEWORK TO DEVELOP FLAW ACCEPTANCE CRITERIA FOR STRUCTURAL INTEGRITY ASSESSMENT OF MULTIPURPOSE CANISTERS FOR EXTENDED STORAGE OF USED NUCLEAR FUEL

A multipurpose canister (MPC) made of austenitic stainless steel is loaded with used nuclear fuel assemblies and is part of the transfer cask system to move the fuel from the spent fuel pool to prepare for storage, and is part of the storage cask system for on-site dry storage. This weld-sealed canister is also expected to be part of the transportation package following storage. The canister may be subject to service-induced degradation especially if exposed to aggressive environments during possible very long-term storage period if the permanent repository is yet to be identified and readied. Stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone because the construction of MPC does not require heat treatment for stress relief. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic Inservice Inspection. The external loading cases include thermal accident scenarios and cask drop conditions with the contribution from the welding residual stresses. The determination of acceptable flaw size is based on the procedure to evaluate flaw stability provided by American Petroleum Institute (API) 579 Fitness-for-Service (Second Edition). The material mechanical and fracture properties for base and weld metals and the stress analysis results are obtained from the open literature such as NUREG-1864. Subcritical crack growth from stress corrosion cracking (SCC), and its impact on inspection intervals and acceptance criteria, is not addressed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1985

A potentiodynamic study of aluminum-lithium alloys in an aqueous sodium chloride environment

The characteristics of the potentiodynamic curves for Al-Li alloys in 3.5 percent NaCl aqueous solution are explained and the electrochemical parameters of the potentiodynamic technique are correlated to observed pitting and intergranular cracking behavior. It is shown that the oxygen content of the sodium chloride electrolyte plays an important role in the electrochemical behavior of Al-Li alloys. The potentiodynamic behavior of the alloys is found to be insensitive to variation in compositional content and heat treatment, both of which affect the stress-corrosion behavior. Stringer oxide particle attack and random pitting are observed. It is shown that alternate-immersion exposure prior to potentiodynamic polarization may offer a means of assessing susceptibility to stress-corrosion cracking.

Tsao, C.-H. T.↗
NASA NTRS2003

Temperature, stress, and corrosive sensing apparatus utilizing harmonic response of magnetically soft sensor element (s)

A temperature sensing apparatus including a sensor element made of a magnetically soft material operatively arranged within a first and second time-varying interrogation magnetic field, the first time-varying magnetic field being generated at a frequency higher than that for the second magnetic field. A receiver, remote from the sensor element, is engaged to measure intensity of electromagnetic emissions from the sensor element to identify a relative maximum amplitude value for each of a plurality of higher-order harmonic frequency amplitudes so measured. A unit then determines a value for temperature (or other parameter of interst) using the relative maximum harmonic amplitude values identified. In other aspects of the invention, the focus is on an apparatus and technique for determining a value for of stress condition of a solid analyte and for determining a value for corrosion, using the relative maximum harmonic amplitude values identified. A magnetically hard element supporting a biasing field adjacent the magnetically soft sensor element can be included.

Grimes, Craig A.↗
DOE OSTI2022

Characterization of the Aerosol Source Term in Dry Storage Canisters

As long-term dry storage of used nuclear fuel at independent spent fuel storage installations (ISFSI) trends toward the de facto back end of the US fuel cycle, it becomes appropriate to investigate potential degradation and dispersion scenarios for suitable risk mitigation purposes. Pitting and subsequent stress corrosion cracking of the canister wall is currently viewed as a potential scenario leading to a through-wall pathway for contamination to be transferred from within the storage container to the surrounding environment. While stress corrosion cracking measurements are currently underway to further characterize this scenario, a parallel effort endeavors to perform a consequence analysis of conditions in which through wall cracks are indeed formed. This effort consists of engineering scale modeling using the GOTHIC and MELCOR software packages along with experimental depletion and penetration tests.

spent fuel storage, aerosol deposition, Stress Cor↗