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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 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
NASA NTRS1994

WINCLR: a Computer Code for Heat Transfer and Clearance Calculation in a Compressor

One of the concerns during inclement weather operation of aircraft in rain and hail storm conditions is the nature and extent of changes in compressor casing clearance. An increase in clearance affects efficiency while a decrease may cause blade rubbing with the casing. The change in clearance is the result of geometrical dimensional changes in the blades, the casing and the rotor due to heat transfer between those parts and the two-phase working fluid. The heat transfer interacts nonlinearly with the performance of the compressor, and, therefore, the determination of clearance changes necessitates a simultaneous determination of change in performance of the compressor. A computer code the WINCLR has been designed for the determination of casing clearance, that is operated interactively with the PURDU-WINCOF I code designed previously for determining the performance of a compressor. A detailed description of the WINCLR code is provided in a companion report. The current report provides details of the code with an illustrative example of application to the case of a multistage compressor. It is found in the example case that under given ingestion and operational conditions, it is possible for a compressor to undergo changes in performance in the front stages and rubbing in the back stages.

Bose, T. K.↗
DOE OSTI2026

Additively Manufactured Carbon Fiber-Reinforced Siliconized Silicon Carbide Composites Using Carbon Fiber-Reinforced Poly-Ether-Ether-Ketone (PEEK) as a Precursor

Herein, we report a method to additively manufacture carbon fiber-reinforced siliconized silicon carbide composites. The process involves the pyrolysis of a 3D-printed carbon fiber-reinforced poly-ether-ether-ketone (PEEK) composite to produce a porous carbon fiber-reinforced carbon matrix composite preform, which is subsequently infiltrated with molten silicon to obtain a carbon fiber-reinforced siliconized silicon carbide composite. A key aspect of the method is limiting polymer melt flow during pyrolysis of PEEK, which is achieved by thermally annealing the 3D-printed carbon fiber-reinforced PEEK preform in air at a temperature below PEEK’s melting temperature. Rheological and differential scanning calorimetry (DSC) measurements demonstrate that the thermal annealing treatment altered the melting behavior of PEEK, while NMR and FTIR measurements provided a mechanistic explanation for the structural changes responsible for the behavior. It was also found that dimensional changes during pyrolysis were anisotropic with greater shrinkage in the stacking direction of the material.

Yoon, Bola [ORNL] (ORCID:0000000260875373)↗
NASA NTRS2020

Design and Development of a Robust Chuck Mechanism for the Mars2020 Coring Drill

The Sampling and Caching Subsystem (SCS) onboard the Mars2020 rover includes a coring drill that can perform coring, abrasion, and regolith collection operations. The coring drill requires a chuck mechanism to allow for the changing of the bits needed to perform sampling. The chuck mechanism must be capable of securely connecting a bit under drilling, rover slip, and pullout loads, but also readily release a bit after application of these loads. The Mars2020 chuck consists of a ball lock mechanism that allows for robust engagement and release of bit assemblies that is preloaded to ensure a secure, stiff connection between the bit assembly and the rest of the corer. To prevent jamming, the chuck incorporates rollers to preload the ball lock. The rollers are mounted onto a flexured cam that minimizes changes in internal mechanism loads that result from thermally induced dimensional changes, further mitigating the potential for jamming. The chuck mechanism is moved between engaged, disengaged, and loose chuck positions by rotating the cam with a gearmotor that incorporates a magnetic detent brake that prevents inadvertent mechanism motion under design loads. Ongoing testing is occurring on the qualification model in a dirty environment while coring and abrading rocks in a Mars representative environment.

Barletta, Anthony↗
NASA NTRS1991

Characterization of the dimensional stability of advanced metallic materials using an optical test bench structure

The dimensional stability of low-density high specific-strength metal-matrix composites (including 30 vol pct SiC(p)/SXA 24-T6 Al, 25 vol pct SiC(p)/6061-T6 Al, 40 vol pct graphite P100 fiber/6061 Al, 50 vol pct graphite P100 fiber/6061 Al, and 40 vol pct P100 graphite fiber/AZ91D Mg composites) and an Al-Li-Mg metal alloy was evaluated using a specially designed five-strut optical test bench structure. The structure had 30 thermocouple locations, one retroreflector, one linear interferometer multilayer insulation, and various strip heaters. It was placed in a 10 exp -7 torr capability vacuum chamber with a laser head positioned at a window port, and a laser interferometer system for collecting dimensional change data. It was found that composite materials have greater 40-C temporal dimensional stability than the AL-Li-Mg alloy. Aluminum-based composites demonstrated better 40-C temporal stability than Mg-based composites.

Hsieh, Cheng↗
DOE OSTI2022

Session 3: Irradiation and Thermal Stesses - NRC Graphite Behavior Model

Graphite anisotropic microstructure Graphite crystal structure and basal planes Response to irradiation and temperature C-axis and a-axis behavior The effect of porosity Nano-and micro-length scale Microscopic to macroscopic response Material property changes Defect length-scale factors Internal stress build-up Irradiation induced dimensional change Temperature effects Stress increases & property changes

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1995

Space environmental effects on spacecraft: LEO materials selection guide, part 1

This document provides performance properties on major spacecraft materials and subsystems that have been exposed to the low-Earth orbit (LEO) space environment. Spacecraft materials include metals, polymers, composites, white and black paints, thermal-control blankets, adhesives, and lubricants. Spacecraft subsystems include optical components, solar cells, and electronics. Information has been compiled from LEO short-term spaceflight experiments (e.g., space shuttle) and from retrieved satellites of longer mission durations (e.g., Long Duration Exposure Facility). Major space environment effects include atomic oxygen (AO), ultraviolet radiation, micrometeoroids and debris, contamination, and particle radiation. The main objective of this document is to provide a decision tool to designers for designing spacecraft and structures. This document identifies the space environments that will affect the performance of materials and components, e.g., thermal-optical property changes of paints due to UV exposures, AO-induced surface erosion of composites, dimensional changes due to thermal cycling, vacuum-induced moisture outgassing, and surface optical changes due to AO/UV exposures. Where appropriate, relationships between the space environment and the attendant material/system effects are identified. Part 1 covers spacecraft design considerations for the space environment; advanced composites; polymers; adhesives; metals; ceramics; protective coatings; and lubricants, greases, and seals.

Silverman, Edward M.↗
NASA NTRS1995

Space environmental effects on spacecraft: LEO materials selection guide, part 2

This document provides performance properties on major spacecraft materials and subsystems that have been exposed to the low-Earth orbit (LEO) space environment. Spacecraft materials include metals, polymers, composites, white and black paints, thermal-control blankets, adhesives, and lubricants. Spacecraft subsystems include optical components, solar cells, and electronics. Information has been compiled from LEO short-term spaceflight experiments (e.g., space shuttle) and from retrieved satellites of longer mission durations (e.g., Long Duration Exposure Facility). Major space environment effects include atomic oxygen (AO), ultraviolet radiation, micrometeoroids and debris, contamination, and particle radiation. The main objective of this document is to provide a decision tool to designers for designing spacecraft and structures. This document identifies the space environments that will affect the performance of materials and components, e.g., thermal-optical property changes of paints due to UV exposures, AO-induced surface erosion of composites, dimensional changes due to thermal cycling, vacuum-induced moisture outgassing, and surface optical changes due to AO/UV exposures. Where appropriate, relationships between the space environment and the attendant material/system effects are identified. Part 2 covers thermal control systems, power systems, optical components, electronic systems, and applications.

Silverman, Edward M.↗
DOE OSTI2023

Phase segregation and miscibility of TiO x nanocomposites in Gd-doped ceria solid electrolyte material

Electro-chemo-mechanical (ECM) coupling refers to mechanical deformation due to electrochemically driven compositional change in a solid. An ECM actuator producing micrometre-size displacements and long-term stability at room temperature was recently reported, comprising a 20 mol% Gd-doped ceria (20GDC), a solid electrolyte membrane, placed between two working bodies made of TiO x /20GDC (Ti-GDC) nanocomposites with Ti concentration of 38 mol%. The volumetric changes originating from oxidation or reduction in the local TiO x units are hypothesized to be the origin of mechanical deformation in the ECM actuator. Studying the Ti concentration-dependent structural changes in the Ti-GDC nanocomposites is therefore required for (i) understanding the mechanism of dimensional changes in the ECM actuator and (ii) maximizing the ECM response. Here, the systematic investigation of the local structure of the Ti and Ce ions in Ti-GDC over a broad range of Ti concentrations using synchrotron X-ray absorption spectroscopy and X-ray diffraction is reported. The main finding is that, depending on the Ti concentration, Ti atoms either form a cerium titanate or segregate into a TiO 2 anatase-like phase. The transition region between these two regimes with Ti(IV) concentration between 19% and 57% contained strongly disordered TiO x units dispersed in 20GDC containing Ce(III) and Ce(IV) and hence rich with oxygen vacancies. As a result, this transition region is proposed to be the most advantageous for developing ECM-active materials.

47 OTHER INSTRUMENTATION↗
DOE OSTI2022

Irradiation-Induced Defect Evolution in Nuclear Graphite

Graphite has historically been used as a moderator material in nuclear reactor designs dating back to the first man-made nuclear reactor to achieve criticality (Chicago Pile 1) in 1942. Additionally, graphite is a candidate material for use in the future envisioned next-generation nuclear reactors (Gen IV); specifically, the molten-salt-cooled (MSR) and very-high-temperature reactor (VHTR) concepts. Gen IV reactor concepts will introduce material challenges as temperature regimes and reactor lifetimes are anticipated to far exceed those of earlier reactors. Irradiation-induced defect evolution is a fundamental response in nuclear graphite subjected to irradiation. These defects directly influence the many property changes of nuclear graphite subjected to displacing radiation; however, a comprehensive explanation for irradiation-induced dimensional change remains elusive. The objectives of this project were focused on the characterization of irradiation-induced defect evolution in nuclear graphite via transmission electron microscopy (TEM). With the use of novel TEM specimen preparation techniques, high-temperature electron-irradiation and characterization of high-temperature neutron-irradiated nuclear graphite, novel fullerene-like defects are shown to be a dominant defect type, especially at higher temperatures. These results contradict the historical models of defect evolution and provide valuable insight into the macroscopically observed property changes in irradiated nuclear graphite.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2002

Preliminary Evaluation of Polyarylate Dielectric Films for Cryogenic Applications

Polymeric materials are used extensively on spacecraft and satellites in electrical power and distribution systems, as thermal blankets and optical surface coatings, as well as mechanical support structures. The reliability of these systems when exposed to the harsh environment of space is very critical to the success of the mission and the safety of the crew in manned-flight ventures. In this work, polyarylate films were evaluated for potential use as capacitor dielectrics and wiring insulation for cryogenic applications. Two grades of the film were characterized in terms of their electrical and mechanical properties before and after exposure to liquid nitrogen (-196 C). The electrical characterization consisted of capacitance and dielectric loss measure Cents in the frequency range of 50 Hz to 100 kHz, and volume and surface resistivities. The mechanical measurements performed included changes in tensile (Young's modulus, elongation-at-break, and tensile strength) and structural properties (dimensional change, weight, and surface morphology). The preliminary results, which indicate good stability of the polymer after exposure to liquid nitrogen, are presented and discussed.

Patterson, Richard L.↗
NASA NTRS1989

KOH concentration effect on the cycle life of nickel-hydrogen cells. Part 4: Results of failure analyses

KOH concentration effects on cycle life of a Ni/H2 cell have been studied by carrying out a cycle life test of ten Ni/H2 boiler plate cells which contain electrolytes of various KOH concentrations. Failure analyses of these cells were carried out after completion of the life test which accumulated up to 40,000 cycles at an 80 percent depth of discharge over a period of 3.7 years. These failure analyses included studies on changes of electrical characteristics of test cells and component analyses after disassembly of the cell. The component analyses included visual inspections, dimensional changes, capacity measurements of nickel electrodes, scanning electron microscopy, BET surface area measurements, and chemical analyses. Results have indicated that failure mode and change in the nickel electrode varied as the concentration was varied, especially, when the concentration was changed from 31 percent or higher to 26 percent or lower.

Lim, H. S.↗
NASA NTRS1990

KOH concentration effect on the cycle life of nickel-hydrogen cells. IV - Results of failure analysis

Potassium hydroxide concentration effects on the cycle life of a Ni/H2 cell have been studied by carrying out a cycle life test on ten Ni/H2 boiler plate cells which contain electrolytes of various KOH concentrations. Failure analyses of these cells were carried out after completion of the life test, which accumulated up to 40,000 cycles at an 80-percent depth of discharge over a period of 3.7 years. These failure analyses included studies on changes of electrical characteristics of test cells, and component analyses after disassembly of the cell. The component analyses included visual inspections, dimensional changes, capacity measurements of nickel electrodes, scanning electron microscopy, surface area measurements, and chemical analyses. Results have indicated that failure mode and change in the nickel electrode varied as the concentration was varied, especially when the concentration was changed from 31 percent or higher to 26 percent or lower.

Lim, H. S.↗
DOE OSTI2024

NON-DESTRUCTIVE POST IRRADIATION EXAMINATION OF FIRST CYCLE ACCIDENT TOLERANT AND ADVANCED ZIRCONIUM ALLOY HIGH BURNUP FUEL RODS

Post irradiation examination (PIE) of accident tolerant fuel concepts and high burnup uranium dioxide advanced zirconium alloy clad fuel contribute to near term nuclear industry goals of enhancing light water reactor safety and economics. Westinghouse is developing chromium coated zirconium alloy cladding as an accident tolerant fuel option for current pressurized water reactors. Examination of high burnup fuel further augments the technical foundation for extending the peak rod average burnup limits beyond the current regulatory limit of 62 MWd/kgU. Seven fuel rods were received at the Oak Ridge National Laboratory Irradiated Fuel Examination Facility. This includes three chromium coated zirconium alloy fuel rods and four high burnup advanced zirconium alloy fuel rods. Six of the seven rods were fueled with standard uranium dioxide while one of the chromium coated rods contained Westinghouse’s ADOPT fuel. The overall hot-cell PIE plan and some PIE results have been presented previously. This paper focuses on the additional non-destructive examination and fission gas release. Highlights from visual examination of these rods are presented. The chromium coating showed little to no visual change following irradiation. Observations of the axial burnup trends in the fuel rods are derived from full rod axial gamma-ray scans with additional observations related to local fission product migration. The axial dimensional changes of the seven rods are evaluated. As expected, creep down had not yet occurred in the chromium coated rods, and the maximum diameter strain in the high burnup advanced zirconium alloy cladding was less than 0.5%. Finally, the steady state fission gas release for these rods was measured ranging from 4.1% to 17.6% which matched expectation from literature. These fuel rods now await further destructive examination.

Harp, Jason↗
DOE OSTI2022

Effect of thermal aging on microstructure and crack growth response of Alloy 152 weld

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40-80 years. Likewise, advanced water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 152 under accelerated thermal aging. The materials studied involved three heats of Alloy 152 used to produce a dissimilar metal weld (DMW) joining an Alloy 690 plate to an Alloy 533 low alloy steel (LAS) plate, thermally aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service). The microstructural characterization by means of synchrotron X-ray did not find evidence of LRO in any of the three heats aged to an equivalent of 60 years of service. Testing in a primary water environment of a heat of Alloy 152 aged at 370°C to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloy. However, the SCC CGR response of the aged sample appears to show a deterioration in performance.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Effect of thermal aging on microstructure and stress corrosion cracking behavior of Alloy 152 weldment

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40-80 years. Likewise, water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 152 under accelerated thermal aging. The materials studied involved three heats of Alloy 152 used to produce a dissimilar metal weld (DMW) joining an Alloy 690 plate to an Alloy 533 low alloy steel (LAS) plate, thermally aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service). The microstructural characterization by means of synchrotron X-ray conducted in small, 0.2 mm - step line scans in the high-deformation regions of the weld root – covering areas spanning from the weld heat affected zone (HAZ) in Alloy 690 to the weld and weld butter on LAS - did not show evidence of LRO in any of the three Alloy 152 heats aged to an equivalent of 60 years of service. Testing in a primary water environment of two heats of Alloy 152 aged at 370°C to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloys. However, the SCC CGR response of the aged samples appears to show a deterioration in performance, confirming our previous observation.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Effect of Thermal Aging on Microstructure and Stress Corrosion Cracking Behavior of an Alloy 152 1st Layer Butter Weldment

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40- 80 years. Likewise, water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is critical for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 152 under accelerated thermal aging. The materials studied involved three heats of Alloy 152 used to produce a dissimilar metal weld (DMW) joining an Alloy 690 plate to an Alloy 533 low alloy steel (LAS) plate, thermally aged at three different temperatures (370°C, 400°C and 450°C) for different durations up to 75,000h (equivalent to 60 years of reactor service). The microstructural characterization by means of synchrotron X-ray conducted in small, 0.2 mm - step line scans in the high-deformation regions of the weld root – covering areas spanning from the weld heat affected zone (HAZ) in Alloy 690 to the weld and weld butter on LAS - did not show evidence of LRO in any of the three Alloy 152 heats aged at 370°C and 450°C to an equivalent of 60 years of service. However, the first weld butter layer has high levels of deformation and is highly susceptible to SCC even in its non-aged condition. Nanohardness testing confirmed the extreme hardening with aging (ΔHV ≅ 100) at two locations within this weldment. In absence of LRO, hardening is suspected to be due to thermally-induced Cr carbide precipitation and coarsening. Testing in a primary water environment of the 1st layer of Alloy 152 weld butter aged at 370°C to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloys. Similarly, the SCC CGR response of the aged weld butter does not appear to show a deterioration in performance, however, the difficult-to-test weldment geometry may affect the test results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Effect of thermal aging and irradiation on microstructure and crack growth response of Alloy 690

Nickel-base Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs reactors are expected to operate for 40-80 years. Likewise, advanced water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 690 during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni 2 Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 690 under accelerated thermal aging and irradiation conditions. The materials studied involved Alloy 690 from two sources: (i) Alloy 690 plate from an Alloy 152 dissimilar metal weld (DMW) joining the Alloy 690 plate to Alloy 533 LAS, aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service), and (ii) specimens neutron-irradiated in the BOR-60 reactor up to 40 dpa. For aged Alloy 690 specimens, hardness was found to increase with aging time, however, the microstructural characterization by means of synchrotron X-ray did not find evidence of LRO. The microstructural characterization of neutron-irradiated specimens by TEM found no evidence of LRO either. Testing in a primary water environment of Alloy 690 specimens aged to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloy. The SCC CGR response was also low. Overall, the two Alloy 690 heats investigated in this work, aged up to 60-year service equivalents or exposed to neutron irradiation up to 40 dpa, did not exhibit a deterioration in microstructure or performance.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗