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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 955 records · Page 53
All sourcesNASA NTRSDOE OSTI
NASA NTRS2018

Effect of Control Mode and Load Rate on Fracture Toughness of Brittle Ceramics

The effects of control mode and rate on the fracture toughness of ceramics were measured by using chevron-notched flexure specimen in accordance with ASTM C1421. The use of stroke control gave consistent results with about 2% variation in measured fracture toughness for a very wide range of rates (0.005 to 0.5 mm/min). Use of strain or CMOD control gave an approximately 5% variation over a very wide range of rates, with the measurements being a function of rate. However, the effect was eliminated by use of dry nitrogen, implying a stress corrosion effect rather than a stability effect. With the use of nitrogen for strain control, fracture toughness values were within about 1% over a wide range of rates (1 to 80 micro-epsilon/sec.). CMOD or strain control did allow stable crack extension well past maximum load, and thus is preferred for energy calculations. The effort is being used to confirm recommendations in ASTM International standard C1421 on fracture toughness measurement.

control mode↗
NASA NTRS2018

Effect of Control Mode and Test Rate on Fracture Toughness of Advanced Ceramics

The effects of control mode and rate on the fracture toughness of ceramics were measured by using chevron-notched flexure specimen in accordance with ASTM C1421. The use of stroke control gave consistent results with about 2% variation in measured fracture toughness for a very wide range of rates (0.005 to 0.5 mm/min). Use of strain or CMOD control gave ~5% variation over a very wide range of rates, with the measurements being a function of rate. However, the effect was eliminated by use of dry nitrogen, implying a stress corrosion effect rather than a stability effect. With the use of nitrogen for strain control, fracture toughness values were within about 1% over a wide range of rates (1 to 80 /s). CMOD or strain control did allow stable crack extension well past maximum load, and thus is preferred for energy calculations. The effort is being used to confirm recommendations in ASTM International standard C1421 on fracture toughness measurement.

fracture toughness↗
NASA NTRS2019

Failure of Nd:YVO4 Amplifier Crystals

Brittle single crystals are used in NASA applications ranging from lenses to centrally heated laser slabs. Despite standard procedures to design such components, unexpected failures occasionally occur. The ICEsat-2 (Ices,Cloud, and Elevation satellite-2) employs Yttrium orthovanadate single crystals in laser amplifiers and oscillators. Although the systems are currently flying and successfully operating, some unexpected crystal fractures occurred just prior to flight. The failures were traced to poor crystal quality along with time-dependent chemical reaction within the system assembly that increased stress beyond expectations and promoted both fast fracture and stress corrosion. This presentation will discuss failure analysis of the crystals and the chemical reactions that promoted failures, along with corrective actions taken to reduce likelihood of fracture over the three-year mission. Re-designed lasers were stored for 12 months and operated for a total of 1,000 hours without signs of degradation.

Design↗
NASA NTRS2019

Failure of Nd:YVO4 Amplifier Crystals

Brittle single crystals are used in NASA applications ranging from lenses to centrally heated laser slabs. Despite standard procedures to design such components, unexpected failures occasionally occur. The ICEsat-2 (Ices, cloud, and elevation satellite) employs Yttrium orthovanadate single crystals in laser amplifiers and oscillators. Although the systems are currently flying and successfully operating, some unexpected crystal fractures occurred just prior to flight. The failures were traced to poor crystal quality along with time depend chemical reaction within the system assembly that increased stress beyond expectations and promoted both fast fracture and stress corrosion. This presentation will discuss failure analysis of the crystals and the chemical reactions that promoted failures, along with corrective actions taken to reduce likelihood of fracture over the three year mission. Re-designed lasers were stored for 12 months and operated for a total of 1,000 hours without signs of degradation.

Salem, Jon↗
DOE OSTI2012

Implementation of ASME Code, Section XI, Code Case N-770, on Alternative Examination Requirements for Class 1 Butt Welds Fabricated with Alloy 82/182

In May 2010, the NRC issued a proposed notice of rulemaking that includes a provision to add a new section to its rules to require licensees to implement ASME 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 the Application of Listed Mitigation Activities, Section XI, Division 1,’’ with 15 conditions. Code Case N-770 contains baseline and inservice inspection (ISI) requirements for unmitigated butt welds fabricated with Alloy 82/182 material and preservice and ISI requirements for mitigated butt welds. The NRC stated that application of ASME Code Case N-770 is necessary because the inspections currently required by the ASME Code, Section XI, were not written to address stress corrosion cracking Alloy 82/182 butt welds, and the safety consequences of inadequate inspections can be significant. The NRC expects to issue the final rule incorporating this code case into its regulations in the spring 2011 time frame. This paper discusses the new examination requirements, the conditions that NRC is imposing , and the major concerns with implementation of the new Code Case.

alloy 82/182↗
DOE OSTI2012

Grain boundary characterization in an X750 alloy

Grain boundary chemistry in an X750 Ni alloy was analyzed by atom probe tomography in an effort to clarify the possible roles of elemental segregation and carbide presence on the stress corrosion cracking behavior of Ni alloys. Two types of cracks are observed: straight cracks along twin boundaries and wavy cracks at general boundaries. It was found that carbides (M23C6 and TiC) are present at both twin and general boundaries, with comparable B and P segregation for all types of grain boundaries. Twin boundaries intercept ?’ precipitates while the general boundaries wave around the ?’ and carbide precipitates. Near a crack tip, oxidation takes place on the periphery of carbide precipitate.

36 MATERIALS SCIENCE↗
DOE OSTI2017

High-Resolution Ultrasound Imaging Using Model-Bases Iterative Reconstruction For Canister Degradation Detection

The inspection of the dry storage canisters that house spent nuclear fuel is an important issue facing the nuclear industry; currently, there are limited options available to provide for even minimal inspections. An issue of concern is stress corrosion cracking (SCC) in austenitic stainless steel canisters. SCC is difficult to predict and exhibits small crack opening displacements on the order of 15 30 m. Nondestructive examination (NDE) of such microscopic cracks is especially challenging, and it may be possible to miss SCC during inspections. The coarse grain microstructure at the heat affected zone reduces the achievable sensitivity of conventional ultrasound techniques. At Oak Ridge National Laboratory, a tomographic approach is under development to improve SCC detection using ultrasound guided waves and model-based iterative reconstruction (MBIR). Ultrasound-guided waves propagate parallel to the physical boundaries of the surface and allow for rapid inspection of a large area from a single probe location. MBIR is a novel, effective probabilistic imaging tool that offers higher precision and better image quality than current reconstruction techniques. This paper analyzes the canister environment, stainless steel microstructure, and SCC characteristics. The end goal is to demonstrate the feasibility of an NDE system based on ultrasonic guided waves and MBIR for canister degradation and to produce radar-like images of the canister surface with significantly improved image quality. The proposed methodology can potentially reduce human radiation exposure, result in lower operational costs, and provide a methodology that can be used to verify canister integrity in-situ during extended storage

Chatzidakis, Stylianos↗
DOE OSTI2019

Examination of Cavity Formation in Cold-Worked Alloy 690

Although stress corrosion cracking (SCC) has not been reported in Alloy 690 components after more than 30 years of pressure water reactor (PWR) service, susceptibility to SCC growth in PWR primary water has been reported by many labs for highly CW Alloy 690 along with isolated observations of crack initiation. Post-test examinations on SCC initiation specimens suggested a key role of creep by inducing the formation and growth of grain boundary cavities at carbide interfaces leading to intergranular cracking. However, it is clear that high levels of cold work can produce grain boundary voids and complicates the quantitative evaluation of damage evolution during testing in high-temperature water. Therefore, it is crucial to understand how to properly examine and classify this type of damage in the pre-test condition and establish its evolution with time. This paper outlines our observational protocols on how to best prepare sample surfaces to avoid inducing artifacts, and to perform subsequent examinations utilizing scanning electron microscopy, transmission electron microscopy and focused ion beam milling to assess the formation and development of grain boundary cavities. Detailed examinations will be reviewed on several commercial Alloy 690 heats in the pre-test CW condition and after long-term initiation testing at constant load in 360°C simulated PWR primary water to effectively document grain boundary cavity formation and growth leading to crack initiation.

Olszta, Matthew J.↗
DOE OSTI2019

FRICTION AND PERMEABILITY EVOLUTION OF FRACUTRES AND FAULTS DURING STATIC REPOSE AND DYNAMIC REACTIVATION

The co-evolution of fault/fracture friction and permeability represent important science/engineering challenges impacting natural and engineered systems. Friction directly controls the characteristics of natural fault slip, including inter-seismic healing, slow creep and earthquake rupture. Permeability evolution during fault slip and repose illuminates our understanding of perturbations to the earthquake-modulated natural hydraulic system and of subsurface engineering in recovering both shale gas and geothermal energy and in the safe subsurface sequestration of CO2. However, mechanisms controlling the evolution of friction and permeability during slow slip, fast rupture and inter-event repose are not clearly understood. For instance, the evolution of stick-slip amplitudes and recurrence, the role of inertia and its influence on frictional stability and the interdependence of these processes remain unclear. Furthermore, mechanisms of permeability evolution during the earthquake cycle (repose through rupture) are also poorly understood since they are influenced in a complex way by chemo-mechanical effects such as elastic/plastic compaction, shear comminution, mechanical dilation, pressure solution and stress corrosion. In this study, we explore fracture friction and permeability evolution during static and dynamic reactivation using both numerical and experimental approaches. This is described in the four chapters of this dissertation.

58 GEOSCIENCES↗
DOE OSTI2019

Irradiation damage and IASCC of printed 316L for use as fuel cladding

Additive manufacturing is an area of increasing interest to the nuclear industry. It offers a number of benefits over traditional manufacturing methods, such as the ability to create complex geometries without the need for welds. In addition to the potential for creating nuclear structural components, additive manufacturing may be used to create novel fuel/cladding designs. However, printed steel behaves differently than traditional wrought material. A number of industries are already studying the effects of the printed microstructures on mechanical properties. For cladding material and some structural nuclear components, however, the nuclear industry must also understand the material response to irradiation. This work examines the irradiation damage and the irradiation assisted stress corrosion cracking behavior of both proton irradiated (1 and 5 dpa) printed 316L steel and wrought 316L steel that were irradiated side by side at the Michigan Ion Beam Laboratory. As orientation with respect to the build direction can affect the mechanical properties of printed metal, specimens were taken from multiple orientations within the printed steel. For the IASCC testing, specimens were strained to 4% in simulated BWR NWC water (288°C, 0.2 µS/cm). TEM, SEM and X-ray microCT were used to examine coupons of the material as well as tensile bars used in the IASCC testing. This paper will show the key findings from this work, namely the IASCC response of printed metal (as compared to wrought), as well as the effects of the printed microstructure (such as the pre-existing voids) on the irradiation damage.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Examination Of IASCC In 304 Stainless Steel Core Shroud From A Commercial Boiling Water Reactor (BWR)

The 304 stainless steel core shroud and the accompanying weld material from a commercial boiling-water reactor (BWR) experienced an estimated dose of ~3.5 dpa neutron irradiation leading to irradiation-assisted stress corrosion cracking (IASCC). The resulting IASCC crack tips are examined with high-resolution analysis techniques, including scanning electron microscopy (SEM), scanning transmission electron microscopy (STEM), and atom probe tomography (APT), with select examples of directly correlated STEM/APT observations. Analytical observations reveal the crack morphology, composition, microstructure, and any associated corrosion-induced or radiation-induced segregation in the local metal grain boundaries ahead of the propagating SCC cracks. The observed oxide composition and structure were in-line with expectations, consisting of a Cr-rich inner oxide and a Fe-rich outer oxide, with the leading crack tip oxide being Cr-rich. In addition, the leading grain boundary exhibited typical radiation-induced segregation signatures that were largely unperturbed by the accompanying corrosion. Finally, the quantitative capabilities and limitations of STEM and APT in the context of corrosion fronts and grain boundary chemistry from the directly correlated measurements are discussed.

Olszta, Matthew J.↗
DOE OSTI2022

Intergranular oxidation and oxygen transport in Ni-20Cr during exposure to model dry and wet oxidation environments

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.

Weiser, Martin↗
DOE OSTI2022

The Hanford Lead Canister Collaboration

The Hanford Site, Management of the Cesium and Strontium Capsules (MCSC) Project is preparing to place cesium and strontium material into dry storage using stainless steel canisters that are housed inside vertical concrete casks. Part of the site’s aging management program is to use a spare canister system as a leading indicator canister to provide advance warning for signs of chloride induced stress corrosion cracking (CISCC) or any other kind of canister degradation that might take place over the system’s 300-year design life. The Hanford Lead Canister (HLC) will use electric heaters to simulate the decay heat of cesium and strontium to ensure the HLC simulates the environment and conditions the capsule-loaded systems will experience. In addition to being part of the site’s aging management program, the HLC is being made available for research and development that advances the industry regarding canister mitigation and repair (M&R). The HLC system closely resembles current spent nuclear fuel (SNF) dry storage systems, making the HLC useful for demonstrating technologies or performing research that is relevant to SNF canisters . Several organizations are collaborating to prepare the HLC, along with technical strategy, to perform the needed research and development activities, to include the Pacific Northwest National Laboratory (PNNL), Central Plateau Cleanup Company (CPCCo), and the Electric Power Research Institute (EPRI). In one example of this collaborative research, EPRI is planning to use the HLC to demonstrate in-situ M&R technologies under controlled conditions before the HLC is deployed at Hanford. There is a period of several years when the HLC will be available for canister M&R activity before it is deployed. This paper describes the development of the HLC, its role in the Hanford site’s MCSC Project aging management program, and some of the research activities that are planned for the HLC.

Klymyshyn, Nicholas A.↗
DOE OSTI2022

In-Cell Thermal Creep Frame Capability: Out-of-Cell Construction and Subsized Sample Testing

This report describes the activities related to the construction of an in-cell thermal creep test facility for small-scale specimens in the irradiation-assisted stress-corrosion cracking (IASCC) hotcell located at the Fuels and Applied Science Building (FASB) at INL. The whole project includes three stages: (i) conceptual design; (ii) out-of-cell demonstration tests; and (iii) in-cell demonstration testing. During the current fiscal year (FY)-2022, materials and equipment required to construct the conceptual design outlined in a previous report were procured. Frames for conducting out-of-cell testing have been constructed, and testing plans have been developed to assess the operation of the newly constructed frames. Preliminary testing of flat subsized tensile specimens on ASTM creep frames has also been completed for two sample thicknesses: 0.75 mm and 1.0 mm.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Results of FY 2023 Alloy 617 and Alloy 709 High-Temperature Crack-Growth Testing

This report summarizes the work performed at Idaho National Laboratory under the “Creep, fatigue and creep-fatigue crack growth tests” task of the “Long-Term VHTR Material Qualification – INL” work package. Work was performed this year to validate the crack growth monitoring setup used with the test frames against both continuous crack length monitoring using an optical camera, as well as post-mortem analysis of marker bands on the fracture surface. Delays prevented the use of a gauge to measure load-line displacement for early tests, and so the creep-fatigue crack growth setup was evaluated using the actuator displacement for a load-line displacement setup. While not ideal, this allowed examination of the shortcomings with the current software setup that was originally designed for performing stress corrosion crack growth rate studies. The method for data collection was modified to link the crack growth monitoring software with Instron’s Wave Matrix software. Once the load-line displacement gauge arrived, the crack growth equipment was successfully modified to permit continuous monitoring of both load-line displacement and crack length. This is critical for creep-fatigue and creep crack growth rate studies for ductile material, as it permits the C* and Ct analyses. Test results are shown for Alloy 617 fatigue and creep-fatigue (without the gauge for load-line displacement measurements), as well as creep-fatigue of Alloy 709, which was performed with the completed test setup, allowing for crack length and load-line displacement monitoring.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Pellet Cladding Interaction In-Reactor Ramp Testing in a World without the Halden Boiling Water Reactor

One of the most crucial performance areas for fuel rods in water cooled nuclear power plants is interaction between cladding tubes and fuel pellets. Experimental programs in test reactors have provided key data to help fuel developers and plant operators understand Pellet Cladding Interaction (PCI) phenomena and optimize their strategies for reliable fuel performance. Approximately 50 years of “ramp” testing programs, where the fission heating rate is deliberately manipulated in test rods, have been performed in a handful of test reactors to reveal and understand PCI behaviors such as iodine-assisted stress corrosion cracking. Unfortunately, the test reactors most engaged in this type of work have all been retired over the years up to the recent and unexpected closure of the Halden Boiling Water Reactor which effectively caused a hiatus in PCI ramp testing programs. The need for ramp testing is crucial at this time to enable refined PCI understanding as more plants consider implementing flexible operations, increased fuel rod burnup limits, and new fuel technologies with enhanced accident tolerance. This paper reviews some of the key PCI phenomena that must be addressed with in-pile testing and surveys past test reactor’s methods for achieving the needed conditions. A strategic approach is then presented using test reactors and complimentary facilities which are still available today. Near term data opportunities are put forth along with capability development strategies that will ensure future longevity in this field of research.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

FLAW STABILITY ANALYSIS OF SURFACE CRACKS IN DOE STANDARD CANISTERS UNDER OPERATION LOADS AND WELDING RESIDUAL STRESSES

There are over 3000 commercial spent nuclear fuel (SNF) storage canisters that are made of stainless steel and planned for multi-purpose functions of SNF handling: interim storage, transportation, and ultimate disposal at a future SNF disposition location. To date, many multi-purpose canisters (MPC) are located in coastal regions for long-term storage. The canisters are fabricated by welding, and post-welding heat treatment is not required for relieving welding residual stresses (WRS). As a result, these canisters may be susceptible to chloride induced stress corrosion cracking (CISCC) due to the chloride-bearing marine salts. The flaw stability analysis of MPC canisters was performed at SRNL (i.e., PVP2016-4935 and PVP2016-63887) for through-wall flaws and for surface flaws respectively, where WRS was considered, and the failure assessment diagram (FAD)-based fracture mechanics method codified by API 579-1/ASME FFS-1-2007 Edition was adopted.

DOE standard canister↗
DOE OSTI2025

Impact of Irradiation on Microstructure and Mechanical Properties of Materials Produced by Advanced Manufacturing

Advanced non-light water reactor designs, known as Generation IV (Gen IV) reactors, typically operate at higher temperatures and under more extreme radiation conditions than conventional light water reactors. A critical aspect of the successful deployment and advancement of Gen IV reactor designs is the selection of appropriate structural materials for specific applications, which necessitates the timely development of new materials and manufacturing processes. Advanced manufacturing (AM) offers numerous opportunities for innovative designs, enabling the production of high-performance components with potentially shorter development cycles compared to traditional manufacturing methods. However, a significant challenge in deploying AM technologies in the nuclear energy sector is the current lack of data on the irradiation performance of AM-produced components. This presentation will discuss neutron and ion irradiation results of AM materials, including stainless steel 316L, Grade 91, SA508, Inconel 718, and Inconel 625 materials. The materials were manufactured by AM, such as Powder Metallurgy Hot Isostatic Pressing (PM HIP), Laser Powder Bed Fusion (LPBF) and Directed Energy Deposition (DED). The effects of neutron irradiation on microstructure (e.g., dislocations, loops, and nanoclusters), tensile properties (e.g., strength and ductility), and ion irradiation on Irradiation-Assisted Stress Corrosion Cracking (IASCC) will be explored. Results are collected from several projects supported by the Nuclear Science User Facilities (NSUF) program.

36 - MATERIALS SCIENCE↗