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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Reduced Diameter and Lined Cladding Production Status

A contract has been placed with Century Tube Inc. to produce cladding for the FAST experiment. Excess EBR II cladding tubes will be cold drawn down to 0.115 in. diameter × 0.009 in. wall and 0.077 in. × 0.006 in. wall. Lined cladding will also be tested in the FAST experiment. Century Tube will produce several samples, varying assembly the assembly process, with solid zirconium or vanadium rods inside of the HT9. Cladding/liner bonding of each will be evaluated using microscopy. The best one will be further processed to form a liner tube within the cladding. A shrink fit approach is also being developed, where a solid zirconium or vanadium bar is shrunk fit into a cladding piece. This process has shown favorable results and will be compared against the co drawn samples when available.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Coherence Weighting Applied to FMC/TFM Data from Austenitic CRA Clad Lined Pipes

Lined pipe girth weld regions are challenging to inspect using ultrasonic techniques due to their anisotropic and inhomogeneous characteristics, which lead to noise and acoustic beam distortion. This paper is focused on the application of coherence weighting methods to decrease the background noise level and false indications on total focusing method (TFM) images, encompassing the triple boundary of mechanically lined pipes. Three coherence weighting approaches were used here; the coherence factor (CF), the sign CF (SCF) and the phase CF (PCF). All three were found capable to improve the TFM images, however, the CF and the SCF performed better due to their exponential sensitivity factors, unlike the PCF which has a linear sensitivity factor.

42 ENGINEERING↗
DOE OSTI2016

Fabrication of rodlets with coated and lined Mo-alloy cladding for testing and irradiation

Molybdenum alloy fuel cladding with an outer protective layer has been proposed to achieve enhanced tolerance to severe loss of coolant accidents for light water reactors. The accident tolerant fuel cladding will need to meet stringent fuel reliability requirements and significantly delay steam reaction during a severe accident to provide plant operators additional coping time to recover from the accident. A laboratory measurable target for this feasibility study with coated Mo cladding is to demonstrate its capability to survive in steam and maintain fuel rods in coolable geometry at 1200-1500 Celsius degrees for at least 24 hours. Preliminary studies indicate such a capability may provide additional coying of 4-19 hours. A feasibility study undertaken since 2012 has demonstrated feasibility of fabricating long, thin-wall Mo and Mo-ODS tubes, and forming a metallurgically bonded protective coating of FeCrAl or Zr-alloy via physical vapor deposition to provide protection of the Mo cladding in 1200 C. degrees for 24 hours. The diametral ductility of the Mo and Mo-ODS thin wall tubes can be increased to 5-15% pre-irradiation to meet fuel operational requirements. Fabrication of the outer protective layer using thermo-mechanical processing, such as hot isostatic pressing and hot coextrusion, has been under evaluation to achieve high quality lined Mo cladding from commercial scale production. Welding of lined Mo tubes to end-caps are being developed using a combination of solid state welding of laser or electron beam seal welding. Rodlets of lined Mo tubes are being fabricated for further testing to validate its reliability under normal operation and capability of accident tolerance to the 1200- 1500 C. degrees range. Test reactor irradiation is also being planned. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

BISON BWR Fuel Modeling Capability: Material Models (FY2020)

Until this year the BISON fuel performance code has been developed primarily for pressurized water reactor (PWR), accident tolerant fuel (ATF), and advanced reactor (e.g., tri-structural isotropic (TRISO), metallic) analyses. The main focus of this work is to develop and document required material models that would aid predicting cladding failure with a specific hydride distribution to be investigated for boiling water reactor (BWR) applications in the next fiscal year. In terms of materials, the primary differences between a BWR and PWR is the use of gadolinia doping of the UO 2 fuel, the use of Zircaloy-2 for the cladding, and the addition of a liner on the inner surface of the cladding for improved pellet-clad mechanical interaction (PCMI)performance. The liner is typically made of pure zirconium or a lower tin content zirconium alloy (e.g. Zr-0.3%Sn). This work details the development and documentation of material and behavioral models for BWR analysis such as the addition of material and behavior models for gadolinia doped UO 2 , pure zirconium, Zr-0.3%Sn, and Zircaloy-2 as well as the ease-of-use additions to allow internal meshing capabilities to include liners. Another important aspect affecting cladding performance is hydrogen diffusion and hydride precipitation in the cladding. A model was added previously for non-lined, homogeneous claddings. In this work, instructions are provided on how to apply the model to BWR liner cladding. Representative experimental values are provided for the hydrogen diffusion parameters and precipitation-dissolution kinetics for typical liner materials of interest.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Topological photonic crystal fiber

Photonic crystal fibers (PCFs) provide a versatile platform for various applications, thanks to the flexibility with which light guiding can be customized by modifying the fiber geometry. We realize a PCF with guided modes produced by photonic band structure topology rather than conventional mode-trapping mechanisms. The design, which is compatible with the stack-and-draw fabrication process, consists of a cross-sectional photonic topological crystalline insulator with a disclination. A bulk-defect correspondence produces degenerate topological modes, lying below the cladding light line. We use various theoretical methods to confirm their topological origins, including a spectral localizer that makes minimal assumptions about the band structure. Our experiments on the fabricated fiber show it transmitting visible to near-infrared light with low losses of 10 to 20 decibels per kilometer, which do not increase substantially when the fiber is bent. A comparable solid-core PCF of conventional design exhibits substantially higher bending losses. Optical fibers based on topological modes hold promise for improved performance and versatile functionalities.

Zhu, Bofeng [Nanyang Technological Univ. (Singapor↗
NASA NTRS1996

Sliding Contact Bearings for Service to 700 C

Cylindrical, sliding contact bearings made entirely of a self-lubricating powder metallurgy composite (PM212) or of super alloy shells lined with clad PM212 were tested in an oscillating mode at temperatures from 25 to 700 C. Tests of 100 hr duration or longer were conducted at a bearing unit load of 3.45 Mpa (500 psi). Shorter duration tests at various unit loads up to 24.1 Mpa (3500 psi) were also conducted. In comparison tests, bearings lubricated with PM212 had superior anti-wear characteristics compared to the baseline, unlubricated, super alloy bearings: no galling of PM212-lubricated bearings occurred, while severe surface damage including galling occurred, especially at high loads, during the baseline tests. A heat treatment procedure, which dimensionally stabilizes PM212 and thereby minimizes clearance changes during high temperature bearing operation, is described.

Sliney, Harold E.↗
DOE OSTI2014

Cook stove assembly

A combustion chamber, having an upper part and a lower part, may include an annular constriction, in combination with the combustion chamber, to aid in directing partially combusted gases such as carbon monoxide away from the periphery of the combustion chamber back toward its center, and into the flame front. The annular constriction may also impede the flow of partially combusted gases located at the periphery, thus increasing the time these gases spend within the combustion chamber and increasing the likelihood that any products of incomplete combustion will undergo combustion. The combustion chamber may further comprise a dual burner cooktop for directing combustion gases and exhaust to multiple cooking vessels. In further embodiments, the combustion chamber may be made of, lined, or clad with a metal alloy comprising iron, chromium, and aluminum.

09 BIOMASS FUELS↗
NASA NTRS1984

Effect of Interfacial characteristics of metal clad polymeric substrates on electrical high frequency interconnection performance

Etched metallic conductor lines on metal clad polymeric substrates are used for electronic component interconnections. Significant signal losses are observed for microstrip conductor lines used for interconnecting high frequency devices. At these frequencies, the electronic signal travels closer to the metal-polymer interface due to the skin effect. Copper-teflon interfaces were characterized by scanning electron microscopy (SEM) and Auger electron spectroscopy (AES) to determine the interfacial properties. Data relating roughness of the copper film to signal losses was compared to theory. Films used to enhance adhesion are found, to contribute to these losses.

Bhasin, K. B.↗
NASA NTRS1975

Design, fabrication, and operation of capsules for the irradiation testing of candidate advanced space reactor fuel pins

Fuel irradiation experiments were designed, built, and operated to test uranium mononitride (UN) fuel clad in tungsten-lined T-111 and uranium dioxide fuel clad in both tungsten-lined T-111 and tungsten-lined Nb-1% Zr. A total of nine fuel pins was irradiated at average cladding temperatures ranging from 931 to 1015 C. The UN experiments, capsules UN-4 and -5, operated for 10,480 and 10,037 hr, respectively, at an average linear heat generation rate of 10 kW/ft. The UO2 experiment, capsule UN-6, operated for 8333 hr at an average linear heat generation rate of approximately 5 kW/ft. Following irradiation, the nine fuel pins were removed from their capsules, externally examined, and sent to the NASA Plum Brook Facility for more detailed postirradiation examination. During visual examination, it was discovered that the cladding of the fuel pin containing dense UN in each of capsules UN-4 and -5 had failed, exposing the UN fuel to the NaK in which the pins were submerged and permitting the release of fission gas from the failed pins. A rough analysis of the fission gas seen in samples of the gas in the fuel pin region indicated fission gas release-to-birth rates from these fuel pins in the range of .00001.

Thoms, K. R.↗
NASA NTRS1974

Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept

An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.

Bowles, K. J.↗
DOE OSTI2016

Uncertainty quantification and sensitivity analysis with CASL Core Simulator VERA-CS

Uncertainty quantification and sensitivity analysis are important for nuclear reactor safety design and analysis. A 2x2 fuel assembly core design was developed and simulated by the Virtual Environment for Reactor Applications, Core Simulator (VERA-CS) coupled neutronics and thermal-hydraulics code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). An approach to uncertainty quantification and sensitivity analysis with VERA-CS was developed and a new toolkit was created to perform uncertainty quantification and sensitivity analysis with fourteen uncertain input parameters. Furthermore, the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel center-line temperature, and maximum outer clad surface temperature were chosen as the selected figures of merit. Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis and coolant inlet temperature was consistently the most influential parameter. We used parameters as inputs to the critical heat flux calculation with the W-3 correlation were shown to be the most influential on the MDNBR, maximum fuel center-line temperature, and maximum outer clad surface temperature.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1974

Evaluation of tantalum-alloy-clad uranium mononitride fuel specimens from 7500-hour, 1040 C pumped-lithium-loop test

Simulated nuclear fuel element specimens, consisting of uranium mononitride (UN) fuel cylinders clad with tungsten-lined T-111, were exposed for up to 7500 hr at 1040 C (1900 F) in a pumped-lithium loop. The lithium flow velocity was 1.5 m/sec (5 ft/sec) in the specimen test section. No evidence of any compatibility problems between the specimens and the flowing lithium was found based on appearance, weight change, chemistry, and metallography. Direct exposure of the UN to the lithium through a simulated cladding crack resulted in some erosion of the UN in the area of the defect. The T-111 cladding was ductile after lithium exposure, but it was sensitive to hydrogen embrittlement during post-test handling.

Watson, G. K.↗
NASA NTRS1975

Materials technology for an advanced space power nuclear reactor concept: Program summary

The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

Gluyas, R. E.↗
DOE OSTI2015

Interfacial microstructure and properties of copper clad steel produced using friction stir welding versus gas metal arc welding

A preliminary study compares the feasibility and microstructures of pure copper claddings produced on a pressure vessel A516 Gr. 70 steel plate, using friction stir welding versus gas metal arc welding. A combination of optical and scanning electron microscopy is used to characterize the grain structures in both the copper cladding and heat affected zone in the steel near the fusion line. The friction stir welding technique produces copper cladding with a grain size of around 25 μm, and no evidence of liquid copper penetration into the steel. The gas metal arc welding of copper cladding exhibits grain sizes over 1 mm, and with surface microcracks as well as penetration of liquid copper up to 50 μm into the steel substrate. Transmission electron microscopy reveals that metallurgical bonding is produced in both processes. Increased diffusion of Mn and Si into the copper cladding occurs when using gas metal arc welding, although some nano-pores were detected in the FSW joint interface. - Highlights: • Cladding of steel with pure copper is possible using either FSW or GMAW. • The FSW yielded a finer grain structure in the copper, with no evidence of cracking. • The FSW joint contains some evidence of nano-pores at the interface of the steel/copper. • Copper cladding by GMAW contained surface cracks attributed to high thermal stresses. • The steel adjacent to the fusion line maintained a hardness value below 248 HV.

36 MATERIALS SCIENCE↗
NASA NTRS2018

Demonstration of Subscale Cermet Fuel Specimen Fabrication Approach Using Spark Plasma Sintering and Diffusion Bonding

Nuclear thermal propulsion (NTP) has the potential to expand the limits of human space exploration by enabling crewed missions to Mars and beyond. The viability of NTP hinges on the development of a robust nuclear fuel material that can perform in the harsh operating environment (> or = 2500K, reactive hydrogen) of a nuclear thermal rocket (NTR) engine. Efforts are ongoing to develop fuel material and to assemble fuel elements that will be stable during the service life of an NTR. Ceramic-metal (cermet) fuels are being actively pursued by NASA Marshall Space Flight Center (MSFC) due to their demonstrated high-temperature stability and hydrogen compatibility. Building on past cermet fuel development research, experiments were conducted to investigate a modern fabrication approach for cermet fuel elements. The experiments used consolidated tungsten (W)-60vol%zirconia (ZrO2) compacts that were formed via spark plasma sintering (SPS). The consolidated compacts were stacked and diffusion bonded to assess the integrity of the bond lines and internal cooling channel cladding. The assessment included hot hydrogen testing of the manufactured surrogate fuel and pure W for 45 minutes at 2500 K in the compact fuel element environmental test (CFEET) system. Performance of bonded W-ZrO2 rods was compared to bonded pure W rods to access bond line integrity and composite stability. Bonded surrogate fuels retained structural integrity throughout testing and incurred minimal mass loss.

Barnes, Marvin W.↗
DOE OSTI2014

Microstructural Characteristics of HIP-bonded Monolithic Nuclear Fuels with a Diffusion Barrier

Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative (GTRI) is developing an advanced monolithic fuel to convert US high performance research reactors to low-enriched uranium. Hot-isostatic-press bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U–Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between fuel meat, cladding, and diffusion barrier, as well as U–10Mo fuel meat and Al–6061 cladding were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are • A typical Zr diffusion barrier of thickness 25 µm • Transverse cross section that exhibits relatively equiaxed grains with an average grain diameter of 10 µm • Chemical banding, in some areas more than 100 µm in length, that is very pronounced in longitudinal (i.e., rolling) direction with Mo concentration varying from 7–13 wt% • Decomposed areas containing plate-shaped low-Mo phase • A typical Zr/cladding interaction layer of thickness 1-2 µm • A visible UZr2 bearing layer of thickness 1-2 µm • Mo-rich precipitates (mainly Mo2Zr, forming a layer in some areas) followed by a Mo-depleted sub-layer between the visible UZr2-bearing layer and the U–Mo matrix • No excessive interaction between cladding and the uncoated fuel edge • Cladding-to-cladding bonding that exhibits no cracks or porosity with second phases high in Mg, Si, and O decorating the bond line. • Some of these attributes might be critical to the irradiation performance of monolithic U-10Mo nuclear fuel. There are several issues or concerns that warrant more detailed study, such as precipitation along cladding-to-cladding bond line, chemical banding, uncovered fuel-zone edge, and interaction layer between U–Mo fuel meat and zirconium. Future post-irradiation examination results will focus, among other things, on identifying in-reactor failure mechanisms and, eventually, directing further fresh fuel characterization efforts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗