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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 469 records · Page 26
All sourcesNASA NTRSDOE OSTI
NASA NTRS1985

Influence of voids on the strength of wrought materials

Three-dimensional voids, which are present in most materials, may be satisfactorily modelled by two-dimensional holes (i.e., cylindrical voids) in sheet metal. In this study, the influence of certain orientations and shapes of voids upon the mechanical properties and fracture behavior of certain ductile materials has been studied. The presence of voids is found to exert a negligible influence on the ultimate tensile strength, owing to plastic flow neutralizing the stress intensification present before yielding occurs. The shape and orientation of the defects, however, are seen to play an important role relative to strain at fracture. The maximum intensified tensile stress criterion which holds for brittle materials is found to apply to ductile materials as well.

Shaw, M. C.↗
NASA NTRS1986

A new readily processable polyimide

As part of an effort to develop tough solvent resistance thermoplastics for potential use as structural resins on aerospace vehicles, a new processable polyimide was evaluated. The synthesis involved the reaction of a new diamine, 1,3-bis 2-(3-aminophenoxy)ethyl ether, with 3,3',4,4'-benzophenonetetracarboxylic dianhydride to form the polyamic acid and subsequent conversion of it to the polyimide. Various physical properties such as thermal stability, solvent resistance, glass transition temperature, crystalline melt temperature, melt viscosity and mechanical properties such as fracture toughness, adhesive, film and composite properties are reported. Of particular interest is the extremely high titanium to titanium tensile shear strength obtained for this polyimide.

Harris, F. W.↗
NASA NTRS1987

A new readily processable polyimide

As part of an effort to develop tough solvent resistance thermoplastics for potential use as structural resins on aerospace vehicles, a new processable polyimide was evaluated. The synthesis involved the reaction of a new diamine, 1,3-bis 2-(3-aminophenoxy)ethyl ether, with 3,3',4,4'-benzophenonetetracarboxylic dianhydride to form the polyamic acid and subsequent conversion of it to the polyimide. Various physical properties such as thermal stability, solvent resistance, glass transition temperature, crystalline melt temperature, melt viscosity and mechanical properties such as fracture toughness, adhesive, film and composite properties are reported. Of particular interest is the extremely high titanium to titanium tensile shear strength obtained for this polyimide.

Harris, F. W.↗
NASA NTRS1989

Hydrogen embrittlement in nickel-hydrogen cells

It was long known that many strong metals can become weakened and brittle as the result of the accumulation of hydrogen within the metal. When the metal is stretched, it does not show normal ductile properties, but fractures prematurely. This problem can occur as the result of a hydrogen evolution reaction such as corrosion or electroplating, or due to hydrogen in the environment at the metal surface. High strength alloys such as steels are especially susceptible to hydrogen embrittlement. Nickel-hydrogen cells commonly use Inconel 718 alloy for the pressure container, and this also is susceptible to hydrogen embrittlement. Metals differ in their susceptibility to embrittlement. Hydrogen embrittlement in nickel-hydrogen cells is analyzed and the reasons why it may or may not occur are discussed. Although Inconel 718 can display hydrogen embrittlement, experience has not identified any problem with nickel-hydrogen cells. No hydrogen embrittlement problem is expected with the 718 alloy pressure container used in nickel-hydrogen cells.

Gross, Sidney↗
NASA NTRS1991

Investigation of the effect of thermal exposure on the mechanical properties of titanium/SiC composites

The objective was to evaluate the influence of thermal exposure, both isothermal and cyclic, on the reaction kinetics, mechanical properties and fracture behavior of Ti-1100 alloy/SiC fiber composites. Previously, it was determined that composites made with TiB2 coated SiC fiber (sigma) reacted at the same rate as SCS-6 fibers. As a result, the thinner surface coating on the Sigma fiber was completely consumed at shorter times than that on the SCS-6. Thermal cycling experiments were conducted on longitudinal and transverse Ti-1100/SCS-6 composites over a temperature range of 150 to 800 C for 500 cycles. The thermal exposures were carried out in air and argon. No appreciable tensile strength degradation was observed for samples cycled in argon, although a strength loss was noted for the samples cycled in air. Fracture surfaces characterization showed brittle matrix failure in regions near the surface and in regions where there was a path for O2 ingress.

Gundel, Douglas B.↗
NASA NTRS1995

Nial and Nial-Based Composites Directionally Solidified by a Containerless Zone Process

A containerless electromagnetically levitated zone (CELZ) process has been used to directionally solidify NiAl and NiAl-based composites. The CELZ processing results in single crystal NiAl (HP-NiAl) having higher purity than commercially pure NiAl grown by a modified Bridgman process (CP-NiAl). The mechanical properties, specifically fracture toughness and creep strength, of the HP-NiAl are superior to binary CP-NiAl and are used as a base-line for comparison with the composite materials subsequently studied. Two-phase composite materials (NiAl-based eutectic alloys) show improvement in room temperature fracture toughness and 1200 to 1400 K creep strength over that of binary HP-NiAl. Metallic phase reinforcements produce the greatest improvement in fracture toughness, while intermetallic reinforcement produces the largest improvement in high temperature strength. Three-phase eutectic alloys and composite materials were identified and directionally solidified with the intent to combine the improvements observed in the two-phase alloys into one alloy. The room temperature fracture toughness and high temperature strength (in air) serve as the basis for comparison between all of the alloys. Finally, the composite materials are discussed in terms of dominant fracture mechanism observed by fractography.

Joslin, Steven M.↗
DOE OSTI2010

Making silicon stronger.

Silicon microfabrication has seen many decades of development, yet the structural reliability of microelectromechanical systems (MEMS) is far from optimized. The fracture strength of Si MEMS is limited by a combination of poor toughness and nanoscale etch-induced defects. A MEMS-based microtensile technique has been used to characterize the fracture strength distributions of both standard and custom microfabrication processes. Recent improvements permit 1000's of test replicates, revealing subtle but important deviations from the commonly assumed 2-parameter Weibull statistical model. Subsequent failure analysis through a combination of microscopy and numerical simulation reveals salient aspects of nanoscale flaw control. Grain boundaries, for example, suffer from preferential attack during etch-release thereby forming failure-critical grain-boundary grooves. We will discuss ongoing efforts to quantify the various factors that affect the strength of polycrystalline silicon, and how weakest-link theory can be used to make worst-case estimates for design.

36 MATERIALS SCIENCE↗
DOE OSTI2012

FRACTURE TOUGHNESS OF 9Cr-1MoV AND THERMALLY AGED ALLOY 617 FOR ADVANCED REACTOR APPLICATIONS

Nickel-base Alloy 617 is being considered as a structural material for application in the secondary heat exchanger of the New Generation Nuclear Plant, a very high temperature gas-cooled reactor. Thermal aging of Alloy 617 plate and welds is being performed with tensile, Charpy impact, and fracture toughness tests conducted at temperatures to 950 C. Results of testing for thermal aging to 5,300 h have been obtained and are presented; varying effects of thermal aging temperature and time on fracture toughness are observed. The 9Cr-1MoV (Grade 91) ferritic steel is a candidate for structural applications in the sodium fast reactor. Fracture toughness testing of unaged Grade 91 steel has been performed to evaluate specimen size effects in preparation for future testing of the material in the thermally aged condition. Results for material in the mill-annealed and heat treated conditions are presented and show that this heat of Grade 91 steel does not indicate a small specimen bias on the fracture toughness Master Curve reference temperature.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Effect of twinning on texture evolution of depleted uranium using a viscoplastic self-consistent model

Ductility and fracture toughness is a major stumbling block in using depleted uranium as a structural material. The ability to correctly model deformation of uranium can be used to create process path methods to improve its structural design ability. The textural evolution of depleted uranium was simulated using a visco-plastic self consistent model and analyzed by comparing pole figures of the simulations and experimental samples. Depleted uranium has the same structure as alpha uranium, which is an orthorhombic phase of uranium. Both deformation slip and twin systems were compared. The VPSC model was chosen to simulate this material because the model encompasses both low-symmetry materials as well as twinning in materials. This is of particular interest since depleted uranium has a high propensity for twinning, which dominates deformation and texture evolution. Simulated results were compared to experimental results to measure the validity of the model. One specific twin system, the {l_brace}176{r_brace}[512] twin, was of specific notice. The VPSC model was used to simulate the influence of this twin on depleted uranium and was compared with a mechanically shocked depleted uranium sample. Under high strain rate shock deformation conditions, the {l_brace}176{r_brace}[512] twin system appears to be a dominant deformation system. By simulating a compression process using the VPSC model with the {l_brace}176{r_brace}[512] twin as the dominant deformation mode, a favorable comparison could be made between the experimental and simulated textures. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Micromechanical tests of ion irradiated materials: Atomistic simulations and experiments

We investigated irradiation effects on Fe-Cr binary alloys by using a nano-indentation combined with a continuous stiffness measurement (CSM) technique. We modeled the nano-indentation test by using a finite element method. We could extract the intrinsic hardness and the yield stress of an irradiation hardened region by using a so-called inverse method. SiC micro-pillars of various sizes were fabricated by mask and inductively coupled plasma etching technique and compressed by using flat punch nano-indentation. Compressive fracture strength showed a clear specimen size effect. Brittle-to-Ductile transition at room temperature was observed as the specimen size decreases. The effect of irradiation on the fracture strength of SiC micro-pillars was evaluated by performing ion irradiation with Si ions. We have performed molecular dynamics simulations of nano-indentation and nano-pillar compression tests. Radiation effect was observed which is found to be due to the interaction of dislocations nucleated by spherical indenter with pre-existing radiation defects (voids). These atomistic simulations are expected to significantly contribute to the investigation of the fundamental deformation mechanism of small scale irradiated materials. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Fuel cladding post-quench LOCA embrittlement: mechanical test relevance

During a loss-of-coolant-accident (LOCA), the fuel cladding is submitted to a high temperature increase in a steam environment, along with an internal pressure increase which would eventually lead to ballooning and burst of the cladding. Steam could then penetrate through the burst opening and lead to inner surface oxidation and secondary-hydriding of the cladding. Post-quench cladding rupture can then occur either in two-sided oxidized area with wall thinning due to ballooning, or in single-sided oxidized area with large hydrogen content. Common tests involved in the evaluation of the post-quench ductility are Ring Compression Tests (RCT), Expansion Due to Compression Test (EDC) or Axial Tensile Tests (AT). Questions arise about these mechanical tests, as for instance, the influence on sample failure of the oxide layer thickness partitioning between inner and outer diameter. A systematic study was carried out to compare several mechanical tests on both single-sided and two-sided Zircaloy-4 oxidized samples. Five different mechanical tests were compared, including RCT: C-shaped ring Compression Test (CCT), Axial Tensile test (AT), Ring Tensile Test with well-defined machined gage section (RTT) and plain Ring Tensile Test (pRTT). All samples were taken from cladding tubes which had been oxidized in steam at 1200 Celsius degrees and then water-quenched. Mechanical tests were performed at room temperature and two to three samples were tested for each oxidation conditions and for every sample geometry to assess the reproducibility. Metallographic examinations were carried out to measure the ZrO{sub 2} and αZr(O) layer thicknesses and the ex-β/α(O) phase fraction. Mean hydrogen content measurements were also performed systematically on each sample. Results indicate that the different mechanical tests have not the same sensitivity to the oxide layer partitioning: RTT and pRTT load at failure seems to depend mainly on the cumulated oxide layer thickness rather than on its partitioning, whereas this parameter has an impact on CCT and AT test results. For all tensile tests, stresses at failure were derived from load-displacement curves. The failure of brittle samples was assumed to be governed by linear elastic fracture mechanics. Using conventional stress intensity factor solution, a value of the material fracture toughness controlling the sample failure was determined from experimental results and discussed. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Testing of LWR Surveillance Materials from the High Fluence UCSB ATR-2 Irradiation

The reactor pressure vessel (RPV) in a light water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely degraded, with the degree of toughness loss dependent on the radiation sensitivity of the materials. Of the many significant issues discussed regarding irradiation effects, the issue considered to have the most impact on the current regulatory process is that associated with effects of neutron irradiation on RPV steels at high fluence, for long irradiation times, and as affected by neutron flux. As also stated in a companion paper by Odette, et. al., the primary objective of the DOE-sponsored Light Water Reactor Sustainability (LWRS) Program RPV task is to develop robust predictions of transition temperature shifts (TTS) at high fluence (φt) to at least 10{sup 20} n/cm{sup 2} (>1 MeV) pertinent to plant operation of some pressurized water reactors (PWR) for 80 full power years. While understanding of copper-enriched precipitates (CRPs) have been fully developed, the discovery and experimental verification of 'late blooming' Mn-Ni-enriched precipitates (MNPs) with little to no copper for nucleation has stimulated research efforts to understand the evolution of these phases. The new and existing databases will be combined to support development of physically based models of TTS for high fluence-low flux (φ < 10 {sup 11}n/cm{sup 2}-s) conditions, beyond the existing surveillance database, to neutron fluences of at least 10{sup 20} n/cm{sup 2} (>1 MeV). The RPV task of the LWRS Program has worked with various organizations to obtain archival surveillance materials from commercial nuclear power plants to allow for comparisons of the irradiation-induced microstructural features from reactor surveillance materials with those from similar alloys irradiated under high flux conditions in test reactors, such as the UCSB ATR-2 experiment conducted at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Irradiation Response of the Ferrite Phase in CF3 Cast Stainless Steel

Extending service of Light Water Reactors (LWRs) to beyond 60 years will demand a high integrity of materials and components in reactors. The accurate evaluation and prediction of materials performance under anticipated operation conditions are of particular importance for ensuring the safe operation of nuclear power plants over their extended lifetime. The duplex structure stainless steels including cast austenitic stainless steels (CASS) and welds of austenitic stainless steels are widely used in LWRs. The CASS alloys are mainly used in LWRs primary pressure boundary and for some limited core structural components. The type 308/309 welds are used for welding the reactor components made of 304 and 316 stainless steels and as a cladding layer on the inner surface of reactor pressure vessel. The core support columns made of CF8 CASS in a B and W design PWR are expected to be exposed to a fluence of between 1x10{sup 21} and 1x10{sup 22} n/cm{sup 2}, E>1.0 MeV, and the austenitic stainless steel welds can potentially receive a fluence up to 5x10{sup 22} n/cm{sup 2}, E>1.0 MeV. In addition to the thermal aging effect, the radiation induced fracture toughness in those duplex structure stainless steel components located in the high neutron fluence regions in LWRs hasn't be fully evaluated. Previous studies showed that the spinodal decomposition in the delta ferrite phase in CASS or type 308/309 welds is the primary embrittlement mechanism, and the G-phase precipitates were also identified but with less impact on the reduction of fracture toughness. It is generally agreed that the irradiation induced fracture toughness decrease in CASS and weld steels saturates after 10 dpa (∼1x10{sup 22} n/cm{sup 2}, E>1.0 MeV). Nevertheless, there are not enough high dose data points to confirm the lower bound in measured fracture toughness as the reported doses are limited to 15 dpa. The work in this paper is to present a more comprehensive study on the microstructural evolution in the ferrite phase in a CF3 CASS, and the results provide a scientific basis for extrapolating the change of fracture toughness to a high dose range.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Nondestructive Evaluation of Alkali-Silica Reaction in High-Strength Concrete for Aging Structures Sustainability

Concrete is a multiphase geo-material, which is utilized extensively in passive supporting and shielding structures at nuclear facilities. Chemo-mechanical degradation of concrete in old structures is a challenge to sustainability of aging commercial light water reactors. Alkali-silica reaction (ASR) is a common slowly evolving chemo-mechanical damage mechanism of concrete, which could take tens of years to reach the threshold stage when visible cracks appear in a concrete structure. In ASR, concrete constituents react with each other, namely the alkalis in the pore solution of hydrated cement paste (mostly K{sup +}OH{sup -} and Na{sup +}OH{sup -}) react with non-crystalline silica (SiO{sub 2}) of the aggregates to form highly hydrophilic gel. At sufficient levels of ambient relative humidity (typically above 85%), imbibition of water causes expansion of the ASR gel, which results in cracking of constrained reinforced concrete and ultimate structural weakening below safety margin levels. Stability and functional performance of concrete structures potentially suffering from chemo-mechanical deterioration due to ASR is evaluated in a multi-stage process involving material evaluation and structural engineering analysis. Existing methods used by industry for concrete material evaluation are based on extrusion of cores from multiple locations in a structure, and destructive laboratory tests on cores. Because each destructive test requires separate core extrusion, the total number of cores taken from the structure at once could be numbering in hundreds. Laboratory tests include petrography (high resolution optical microscopy of core sections) to develop damage rating index (DRI) score of concrete cores. Another test may involve chemical pore solution extraction from cores, and solution speciation analysis using inductively coupled plasma and mass spectroscopy (ICP/MS). Other tests may consist of mechanical evaluation of cores strength and fracture toughness. The DRI score of concrete material as well as other information obtained from the laboratory tests is subsequently used in structural engineering calculations, where stresses in the entire structure are modeled to evaluate its safety and mechanical stability. Extrusion of a large number of cores is highly undesirable for evaluation of nuclear facility concrete structures. The amount of destructive material testing could be minimized by developing methods for in-situ NDE of ASR to quantify the amount of deterioration that is occurring in the structure, and to identify areas in the structure where more deterioration may be occurring.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Fracture Behavior Considerations for the TRISO Particle Matrix

To assess whether matrix fracture would result in an unacceptable loss of containment or confinement in TRISO fuel particles, it is crucial to evaluate the micro-tensile strength, fracture toughness, and irradiation effects on matrix materials. Current data must be comprehensive and validated for modeling fractures under various conditions. Relevant material properties surrounding matrix fracture will be discussed during the presentation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗